2024-03-29T01:03:59Z
https://nucet.pensoft.net/oai.php
10.3897/nucet.4.29453
2018-09-25
nucet
JSC “SSC RF-IPPE named after A.I. Leypunsky”, Obninsk, Russia
author
Alekseev, Pavel
JSC “SSC RF-IPPE named after A.I. Leypunsky”, Obninsk, Russia
author
Krotov, Aleksei
JSC “SSC RF-IPPE named after A.I. Leypunsky”, Obninsk, Russia
author
Ovcharenko, Mikhail
JSC “SSC RF-IPPE named after A.I. Leypunsky”, Obninsk, Russia
author
Linnik, Vladimir
2018-09-25
2018-09-25
2018
Nuclear Energy and Technology
2452-3038
4
1
7-11
2018
PA
Alekseev
author
2011
2011
PA
Alekseev
author
2012
2012
GG
Bartolomei
author
1989
Bases theory and methods for computation nuclear power reactor.
1989
512 pp
1994
1994
ENDF/B-VI Data for MCNP TM (1994) LA-12891-M
LA
Gladkov
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2010
Genetic algorism.
2010
368 pp
10.1016/j.ress.2005.11.018
Autonomus termonuclear nuclear power unit for oil and gas structures. Izvestiya vuzov.
AD
Krotov
author
2011
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2011
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10.1007/s10512-009-9145-y
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Kuznetsov
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1977
Nuclear reactors of space nuclear power unit.
1977
240 pp
GE
Lazarenko
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2006
VA
Linnik
author
2005
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2005
70 pp
Electrisity generation system small nuclear power plant using high effective low temperature thermionic process. Izvestiya vuzov.
PA
Maslov
author
2011
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Yadernaya energetika
2011
2
24
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1997
MCNP – General Monte Carlo N-Particle Transport code (1997) LA-12625-M, Vers. 4B
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Polous
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2013
Modern computational technologies for justification the characteristics of nuclear power propulsion systems in design works of creation a new generation of thermionic space nuclear power unit.
2013
26 pp
Program complex for three-dimensional numerical calculation of thermal end electricity property for multiple-elements fuel elements of thermionic NPU. Izvestiya vuzov.
MA
Polous
author
2012
text
Yadernaya energetika
2012
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WF
Sacco
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2016
Small thermionic nuclear power plant 10/100 kWtel (2016) http://www.ippe.ru/innov/1/in1-7.php [in Russian]
VI
Yarygin
author
2016
Space and planetary nuclear power unit with direct energy conversion.
2016
364 pp
AN
Zabud’ko
author
2004
Comparision and analysis parameter of thermionic conversion reactor for space NPU: IPPE Preprint-3025.
2004
28 pp
10.3897/nucet.4.29453
https://nucet.pensoft.net/article/29453/
https://nucet.pensoft.net/article/29453/download/pdf/
https://nucet.pensoft.net/article/29453/download/xml/
The paper investigates the possibility for reducing the radial power peaking factor kr inside the core of a water-cooled water-moderated thermionic converter reactor (TCR). Due to a highly nonuniform power density, the TCR generates less electric power and the temperature increases in components of the thermionic fuel elements, leading so to a shorter reactor life.
A TCR with an intermediate neutron spectrum has its thermionic fuel elements (TFE) arranged inside the core in concentric circles, this providing for a nonuniform TFE spacing and reduces kr. The water-cooled water-moderated TCR under consideration has a much larger number of TFEs arranged in a hexagonal lattice with a uniform pitch. Power density flattening in a core with a uniform-pitch lattice can be achieved, e.g., through using different fuel enrichment in core or using additional in-core structures. The former requires different TFE types to be taken into account and developed while the latter may cause degradation of the reactor neutronic parameters; all this will affect the design’s economic efficiency.
It is proposed that the core should be split into sections with each section having its own uniform lattice pitch which increases in the direction from the center to the periphery leading so to the radial power density factor decreasing to 1.06. The number of the sections the core is split into depends on the lattice pitch, the TFE type and size, the reflector thickness, and the reactor design constraints. The best lattice spacing options for each section can be selected using the procedure based on a genetic algorithm technology which allows finding solutions that satisfy to a number of conditions.
This approach does not require the reactor dimensions to be increased, different TFE types to be taken into account and developed, or extra structures to be installed at the core center.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Thermionic converter reactor
thermionic fuel element
power peaking factor
lattice pitch
genetic algorithm
optimization
Minimize fission power peaking factor in radial direction of water-cooled and water-moderated thermionic conversion reactor core
Research Article
10.3897/nucet.4.28724
2018-09-25
nucet
JSC “State Scientific Center – Research Institute of Atomic Reactors”, Dimitrovgrad, Russia
author
Osipova, Tatiana
JSC “State Scientific Center – Research Institute of Atomic Reactors”, Dimitrovgrad, Russia
author
Starkov, Vladimir
JSC “State Scientific Center – Research Institute of Atomic Reactors”, Dimitrovgrad, Russia
author
Uzikov, Vitaly
https://orcid.org/0000-0003-3946-8666
2018-09-25
2018-09-25
2018
Nuclear Energy and Technology
2452-3038
4
1
1-6
2018
N
Arkhangelsky
author
2013
2013
A
Bychkov
author
2009
2009
VS
Chirkin
author
1968
1968
CD
Fletcher
author
1995
1995
GD
Gataullina
author
2012
2012
PL
Kirillov
author
1990
Reference Book on Thermohydraulic Calculations (Nuclear reactors, Heat-exchange Units, Steam-generating Units).
1990
360 pp
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Kutateladze
author
1990
Heat Transfer and Hydrodynamic Resistance: A Reference Book.
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367 pp
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Mikheev
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1977
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344 pp
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Osipova
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Osipova
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Osipova
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Osipova
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TA
Osipova
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TA
Osipova
author
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Samsonov
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1991
248 pp
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Seredkin
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Methodical Study of Stress Corrosion Cracking of Inconel 718 Alloy in the SM-3 Reactor. Report on the basic research works carried out in 2013.
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31 pp
VA
Tsykanov
author
1973
Materials Irradiation Ttechnique in Rreactors with a High Neutron Flux.
1973
264 pp
VA
Uzikov
author
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2015
AI
Zvir
author
2009
2009
10.3897/nucet.4.28724
https://nucet.pensoft.net/article/28724/
https://nucet.pensoft.net/article/28724/download/pdf/
https://nucet.pensoft.net/article/28724/download/xml/
It has been shown by calculations that it is possible to extend considerably the capabilities for control of temperature conditions in an ampoule channel with natural coolant circulation, using the proposed hydraulic circuit layout, on samples during irradiation in the SM-3 reactor reflector cell by changing the circulation circuit geometry through the arrangement of a bypass heat removal line formed in the upper part above the flow limiter as compared to control only by changing the thermal conductivity of the gas gap in the channel body (through changing the gas pressure or composition). The ampoule channel test conditions, layout and simulation model for thermal-hydraulic analysis using the RELAP5/MOD3.2 code are presented. An investigation was conducted to study the effects of the bypass cooling circuit on the temperature conditions during irradiation of samples in an ampoule channel. The bypass flow rate change is achieved by varying the passage area of the flow limiter orifice. Options have been considered for filling the channel body gas gap with helium and a helium mixture. The calculation showed that the heat removed by the bypass line could reach 40% of the total heat released in the channel. With helium used in the channel body gap, the temperature conditions during irradiation are adjusted in a broader range (200–330 °С) than with a gas mixture of a lower thermal conductivity (279–330 °С), the major temperature variation taking place with the flow limiter orifice area being less than 0.2–0.3 cm2. Any further increase in the flow limiter orifice area does not lead to a major temperature change in the coolant flowing about the samples.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
SM research reactor
natural-circulation ampoule channel
investigation results
irradiation temperature conditions
power density.
Regulation of the temperature in the ampoule channel with natural circulation of coolant
Research Article
10.3897/nucet.4.29838
2018-10-17
nucet
JSC “State Scientific Center – Research Institute of Atomic Reactors”, Dimitrovgrad, Russia
author
Dreganov, Oleg
JSC “State Scientific Center – Research Institute of Atomic Reactors”, Dimitrovgrad, Russia
author
Shulimov, Vitalij
JSC “State Scientific Center – Research Institute of Atomic Reactors”, Dimitrovgrad, Russia
author
Kiselyova, Irina
JSC “State Scientific Center – Research Institute of Atomic Reactors”, Dimitrovgrad, Russia
author
Alekseev, Aleksandr
2018-10-17
2018-10-17
2018
Nuclear Energy and Technology
2452-3038
4
1
21-26
2018
Methods to test VVER fuel in the MIR reactor under transient and accidental conditions. Izvestiya vuzov.
AV
Alekseev
author
2007a
text
Yadernaya Energetika
2007a
3
83
91
Programs and Methods for Testing in the MIR research Reactor Fuel Elements of Water-Cooled Reactors under Conditions Simulating Transient and Emergency Ones.
AV
Alekseev
author
2012
text
Atomnaya Energiya
2012
113
3
146
150
AV
Alekseev
author
2016
2016
Study of VVER-1000 Fuel Rod Behavior under LOCA Conditions. RIAR Proceedings.
AV
Alekseev
author
2017a
text
JSC «SSC RIAR», Dimitrovgrad
2017a
1
12
20
AV
Alekseev
author
2017b
2017b
AV
Alekseev
author
2009
2009
10.1007/s10512-007-0135-7
Studying of the VVER fuel behavior under the control rod ejection accident conditions. In-pile test method and procedure.
AV
Alekseev
author
2006a
text
RIAR Proceedings
2006a
1
23
32
10.1007/s10512-006-0185-2
P
Askeljung
author
2012
2012
AA
Goncharov
author
2016
2016
Integral In-Pile Experiments with VVER-440 and VVER-1000 Multielement Fuel Test Rigs under the LOCA Conditions. Izvestiya vuzov.
AV
Goryachev
author
2004a
text
Yadernaya Energetika
2004a
3
50
58
Integral In-Pile Experiments with VVER-440 and VVER-1000 Multielement Fuel Test Rigs under the LOCA Conditions. Compilation of Experimental Data from a series of Tests. VANT. Ser.
AV
Goryachev
author
2004
text
Fisika Yadernykh Reaktorov
2004
2
29
38
AL
Izhutov
author
2015
2015
Laura
Kekkonen
author
2008
2008
1991
1991
Research Reactors of RIAR and their Experimental Capabilities under Scientific Editorship of Prof. V.A. Tsykanov (1991) NIIAR Publ., Dimitrovgrad, 103 pp. [in Russian]
Experimental and Computational Studies of VVER Fuel Behavior under LOCA Conditions (MIR-LOCA/60 Test). VANT. Ser.
AV
Salatov
author
2013
text
Materialovedenie i Novye Materialy
2013
1
74
26
38
VN
Shulimov
author
2015
2015
VN
Shulimov
author
2004
2004
VP
Spasskov
author
1998
1998
10.3897/nucet.4.29838
https://nucet.pensoft.net/article/29838/
https://nucet.pensoft.net/article/29838/download/pdf/
https://nucet.pensoft.net/article/29838/download/xml/
This paper deals with the problem of measuring the VVER-1000 burnup fuel cladding temperature in a 500–900°C range in the process of experiments in a channel of the MIR research reactor to obtain data on the fuel element behavior under the influence of the parameters typical of the maximum design-basis loss-of-coolant accident (LOCA). Studying the burnup fuel cladding deformation pattern requires measurements of the cladding temperature with no (thermal, mechanical and other) impacts on the cladding in the maximum deformation region.
For dynamic experiments in the MIR reactor channel with fuel testing in a vapor-gas environment, a cladding temperature measuring unit has been developed, in which the cladding is not subjected to external impacts in the maximum deformation region. In the process of being installed into the spacer grid, the thermoelectric transducer (TET) has its hot junction forced against the cladding making it possible to prevent the external impact on the cladding. The thermometric characteristic of the TET attachment, which is associated with the impact of the grid as such on its thermal condition, was studied using a laboratory facility. This technique was used in an in-pile experiment to study the fuel cladding deformation pattern.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Laboratory facility (LF)
experimental fuel element (EFE)
electroheated fuel element simulator (EHFES)
cladding
spacer grid (SG)
thermoelectric transducer (TET)
hot junction
temperature
heat-up rate
MIR reactor
loss-of-coolant accident (LOCA).
Measurement of the spent fuel rod cladding temperature during the in-pile testing at 500 – 900°C
Research Article
10.3897/nucet.4.29842
2018-10-17
nucet
JSC “SSC RF-IPPE n.a. A.I. Leypunsky”, Obninsk, Russia
author
Usanov, Vladimir
JSC “SSC RF-IPPE n.a. A.I. Leypunsky”, Obninsk, Russia
author
Kviatkovskii, Stepan
NRNU “MEPhI”, Moscow, Russia
author
Andrianov, Andrey
2018-10-17
2018-10-17
2018
Nuclear Energy and Technology
2452-3038
4
1
27-33
2018
Towards sustainable nuclear power development.
A
Andrianov
author
2014
text
ATW: International Journal for Nuclear Power
2014
59
5
287
293
AA
Andrianov
author
2017
2017
AF
Egorov
author
2012
2012
AF
Egorov
author
2013
2013
2015
2015
GEN-IV INTERNATIONAL FORUM (2015) Annual Report: 135.
2008
2008
IAEA (2008) Guidance for the Application of an Assessment Methodology for Innovative Nuclear Energy Systems. INPRO Manual. IAEA-TECDOC-1575, Rev. 1, Vienna. http://www-pub.iaea.org/MTCD/Publications/PDF/TE_1575_web.pdf
Overview of Generation IV (Gen IV) Reactor Designs.
2012
text
Safety and Radiological Protection Considerations
2012
2012
13
15
V
Kagramanyan
author
2015
2015
System and economic optimization problems of NNPs and its ideology. Izvestiya vuzov.
AV
Klimenko
author
2016
text
Yadernaya Energetika
2016
1
149
157
V
Kuznetsov
author
2014
2014
10.3390/en8053679
S
Kviatkovskii
author
2017
2017
J
Le Mer
author
2013
2013
2014
2014
OECD (2014) Technology Roadmap Update for Generation IV Nuclear Energy Systems 2014: 14.
N
Ponomarev-Stepnoy
author
2016
The two-component nuclear energy system with thermal and fast reactors in the closed nuclear fuel cycle Ed. by of RAS Acad. N. Ponomarev-Stepnoy.
2016
253 pp
10.1155/2017/9029406
10.3390/su9091623
SM
Goldberg
author
2011
2011
2001
Ministry of the Russian Federation for Atomic Energy. General editor E.O.
2001
128 pp
White Book of Nuclear Power (2001) Ministry of the Russian Federation for Atomic Energy. General editor E.O.Adamov, Moscow, 128 pp. [in Russian]
10.3897/nucet.4.29842
https://nucet.pensoft.net/article/29842/
https://nucet.pensoft.net/article/29842/download/pdf/
https://nucet.pensoft.net/article/29842/download/xml/
The paper describes the approach to the assessment of nuclear energy systems based on the integral indicator characterizing the level of their sustainability and results of comparative assessment of several nuclear energy system options incorporating different combinations of nuclear reactors and nuclear fuel cycle facilities. The nuclear energy systems are characterized by achievement of certain key events pertaining to the following six subject areas: economic performance, safety, availability of resources, waste handling, non-proliferation and public support. Achievement of certain key events is examined within the time interval until 2100, while the key events per se are assessed according to their contribution in the achievement of sustainable development goals. It was demonstrated that nuclear energy systems based on the once-through nuclear fuel cycle with thermal reactors and uranium oxide fuel do not score high according to the integral sustainable development indicator even in the case when the issue of isolation of spent nuclear fuel in geological formation is resolved. Gradual replacement of part of thermal reactors with fast reactors and closing the nuclear fuel cycle results in the achievement of evaluated characteristics in many subject areas, which are close to maximum requirements of sustainable development, and in the significant enhancement of the sustainability indicator.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Nuclear energy system
sustainable development
closed nuclear fuel cycle
fast reactors.
Elaboration of approach to nuclear energy systems assessment by criterion of sustainable development
Research Article
10.3897/nucet.4.29850
2018-10-17
nucet
VNIIAES, Moscow, Russia
author
Baranenko, Valery
VNIIAES, Obninsk, Russia
author
Gulina, Olga
Obninsk Institute for Nuclear Power Engineering, Obninsk, Russia
author
Salnikov, Nikolaj
2018-10-17
2018-10-17
2018
Nuclear Energy and Technology
2452-3038
4
1
35-42
2018
A general diagnostic feature for evaluating the wear rate of pipelines in the weld-affected zones (field experience at the Zaporozhskaya NPP] Zavodskaya laboratoriya.
VI
Baranenko
author
1998
text
Diagnostika Materialov
1998
64
2
56
58
Substantiation of FAC rate and service life estimation under operation control data. Izvestiya vuzov.
VI
Baranenko
author
2016
text
Yadernaya Energetika
2016
2016
2
55
65
On the calculation of FAC rate o and the residual life of NPP. Izvestiya vuzov.
VI
Baranenko
author
2010
text
Yadernaya Energetika
2010
2010
2
55
63
VI
Baranenko
author
2009
2009
2003
2003
Case of ASME Requirements for Analytical Evaluation of Pipe Wall Thinning (2003) Section XI, Division 1. Case N-597-2, 13 pp.
Development of normative documentation for resource management of NPP equipment under FAC conditions.
OM
Gulina
author
2013
text
Yadernaya fizika i inzhiniring
2013
2013
3
273
278
R
Korhonen
author
1994
1994
VA
Nakhalov
author
1983
Reliability of pipe bends in heat power plants.
1983
183 pp
2012
Norms of admissible wall thickness for carbon steel piping under flow accelerated corrosion.
2012
210 pp
RD EO 1.1.2.11.0571-2015 (2012) Norms of admissible wall thickness for carbon steel piping under flow accelerated corrosion.VNIIAES Publ., Мoscow, 210 pp. [in Russian]
2013
EPRI/ 3002000563.
2013
94 pp
Recommendation for Effective Flow-Accelerated Corrosion Program (NSAC-202L-R4) (2013) EPRI/ 3002000563.Technical Report, EPRI, 94 pp.
M
Rushchak
author
1996
1996
1989
Details and assembly units of pearlitic steels for NPP pipelines D = 16–720 mm. Types, construction and dimensions.
1989
155 pp
ОSТ 24.125.30-89 – ОSТ 24.125.57-89 (1989) Details and assembly units of pearlitic steels for NPP pipelines D = 16–720 mm. Types, construction and dimensions.Gosstandart Publ., Leningrad, 155 pp. [in Russian]
1975
Geometric characteristics of bends. Drowning of the stretched part of the bends. Limit deviations in the thickness of the walls of pipelines. Moscow.
1975
8 pp
ОSТ 24.321.26-74, ОSТ 24.321.28-74, ТU 14-3-460-75 (1975) Geometric characteristics of bends. Drowning of the stretched part of the bends. Limit deviations in the thickness of the walls of pipelines. Moscow.Gosstandart Publ., Moscow, 8 pp. [in Russian]
10.3897/nucet.4.29850
https://nucet.pensoft.net/article/29850/
https://nucet.pensoft.net/article/29850/download/pdf/
https://nucet.pensoft.net/article/29850/download/xml/
As of today, large volumes of data related to non-destructive operational control are accumulated on NPPs. For ensuring safe operation of power units, optimization of scope and scheduling operational control it is necessary to continue development of guidance documents, software products, methodological guidance and operational documentation (Baranenko et al. 1998, Gulina et al. 2013, Recommendation (NSAC-202L-R4) 2013).
Approaches are examined to assessment of the rate of erosion-corrosion wear (flow-accelerated corrosion - FAC) according to the data of operational control. The present study was performed based on the data of thickness gauging of different elements of pipelines of NPPs with different types of reactor. Further development of ideas exposed in (Baranenko et al. 2016) allowed revealing specific features of ECW processes on straight sections, bends and in the zones adjacent to weld joints of pipelines of NPPs equipped with VVER and RBMK reactors. Presence of the process of deposition of corrosion products on internal surfaces of pipeline walls results in the fact that residual lifetime of elements nominally increases due to deposition. However, real wall thickness under the layer of deposits is unknown just as the initial wall thickness is unknown as well. Investigation implemented in the present study is aimed at the substantiation of the methodology of calculation of FAC rate according to the data of operational control for the purpose of drawing calculation results closer to the reality keeping conservatism. Uniform approach to the assessment of FAC rate in the examined elements of pipelines was developed. Methodologies for evaluation of correction coefficients taking into account dimensional technological tolerances, special features of geometry of the element, as well as effect of deposits on the results of thickness measurements were suggested based on the data of operational control and industry standards.
The implemented studies demonstrated efficiency of the developed procedures for pipeline welding zones. Analysis of known and newly developed procedures was performed for bends and ranking of these procedures according to the criterion of “conservatism of evaluation of residual lifetime” was executed.
Introduction of correction coefficients allows enhancing conservatism of calculations of lifetime characteristics as compared with calculations performed on the basis of nominal values of thicknesses; the result depends on the type and dimensions of the element, its geometry, as well as on the type of reactor.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Flow-accelerated corrosion (FAC)
thickness gauging
evaluation of FAC rate
bends
welding
residual lifetime.
Flow-accelerated corrosion rate and residual life time estimation for the components of pipeline systems at nuclear power plants based on control data
Research Article
10.3897/nucet.4.29853
2018-10-17
nucet
JSC “Rusatom Automated control systems”, Moscow, Russia
author
Galiev, Ilnar
JSC “Rusatom Automated control systems”, Moscow, Russia
author
Chernyaev, Alexey
JSC “Rusatom Automated control systems”, Moscow, Russia
author
Bibik, Stanislav
2018-10-17
2018-10-17
2018
Nuclear Energy and Technology
2452-3038
4
1
43-50
2018
Standard TETRA. Possibilities and Advantages.
SV
Chivilev
author
2009
text
Elektrotehnicheskie i informacionnye kompleksy i sistemy
2009
5
2
50
55
VA
Gorishnij
author
2009
Assessment of the engineering environment in an emergency situation.
2009
83 pp
J
Hano
author
2011
2011
10.1126/science.228.4702.987
DN
Ljasin
author
2005
Methods and means of protection of computer information.
2005
127 pp
AA
Novikova
author
2013
The natural disaster in Japan and its consequences.
2013
40 pp
2015
2015
NP-001-15 (2015) General provisions for ensuring the safety of nuclear power plants. Rostehnadzor Publ., Moscow. [in Russian]
2001
The design standards of earthquake-resistant nuclear power plants.
2001
33 pp
NP-031-01 (2001) The design standards of earthquake-resistant nuclear power plants.Gosatomnadzor Rossii Publ., Moscow, 33 pp. [in Russian]
E
Obukhov
author
2017
2017
AV
Roslyakov
author
2006
Virtual private networks. Fundamentals of construction and application.
2006
304 pp
V
Shustov
author
2007
Тesting of a new line of seismic base isolators.
2007
35 pp
VS
Sjuvatkin
author
2005
2005
2013
2013
SP 151.13330.2012 (2013) Engineering surveys for the location, design and construction of nuclear power plants. Part I. engineering surveys for the development of pre-project documentation (choice of point and choice of NPP site). Gosstroj Publ., Moscow. [in Russian]
2014
2014
SSG-30 (2014) Safety classification of structures, systems and components in nuclear power plants. Vienna, IAEA.
VO
Tihvinskij
author
2010
LTE Mobile Networks: Technology and Architecture.
2010
284 pp
10.3897/nucet.4.29853
https://nucet.pensoft.net/article/29853/
https://nucet.pensoft.net/article/29853/download/pdf/
https://nucet.pensoft.net/article/29853/download/xml/
Development of seismic protection system for design extension conditions (SPS DEC) is suggested for enhancing safety of operation of NPP located on the territories with unfavorable seismic conditions. The idea of the system consists of the creation of a network of seismic stations arranged at a certain distance from the NPP and equipped with data transmission system. In case of detection by seismic sensors of movements of the ground with magnitude of vibrations exceeding a certain preset value, seismic stations transmit over radio-channel a signal indicating exceedance of the setting before the seismic wave reaches the NPP. This allows initiating transition of the reactor to subcritical operation mode prior to the beginning of destruction of equipment and reactor building. Ensuring reliable protected communication is achieved by simultaneous use of three radio-channels arranged in accordance with TETRA, WiMAx and LTE standards, as well as by the application of appropriate cryptography, authentication and data protection methods for preventing data corruption. Analysis was performed for determining optimal distance between seismic stations and the NPP and optimal number of these stations was determined for the following two options of arrangement of seismic stations: radial arrangement surrounding the NPP and arrangement along the direction towards the place with the highest probability of earthquake incipience. Layout was suggested of multilevel majoritarian data processing for excluding false triggering of the system. Conclusions were formulated on the enhancement of safety of NPP operation and significant reduction of probability of emergency situations due to the generation of anticipatory signals to reactor shutdown systems in case of earthquakes with intensity exceeding maximum design earthquake.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Seismic protection system
maximum design earthquake
reactor shutdown
seismic station
safety
VVER
data transmission.
Development of seismic protection system for design extension conditions
Research Article
10.3897/nucet.4.29855
2018-10-17
nucet
JSC “Pilot and Demonstration Center for Decommissioning of Uranium-Graphite Nuclear Reactors”, Seversk, Russia
author
Bespala, Evgeny
JSC “Pilot and Demonstration Center for Decommissioning of Uranium-Graphite Nuclear Reactors”, Seversk, Russia
author
Pavliuk, Alexander
https://orcid.org/0000-0001-8897-6910
JSC “Pilot and Demonstration Center for Decommissioning of Uranium-Graphite Nuclear Reactors”, Seversk, Russia
author
Zagumennov, Vladimir
JSC “Pilot and Demonstration Center for Decommissioning of Uranium-Graphite Nuclear Reactors”, Seversk, Russia
author
Kotlyarevskiy, Sergey
2018-10-17
2018-10-17
2018
Nuclear Energy and Technology
2452-3038
4
1
51-56
2018
VM
Anischik
author
2003
The modification of instrumental materials by ion and plasma beams.
2003
191 pp
Computer modeling of processes with actinides in radioactive graphite at heating in a nitrogen atmosphere.
NM
Barbin
author
2015
text
Prikladnaya fizika
2015
2015
42
42
47
10.1051/matecconf/20167201011
AV
Bushuev
author
2015
The radioactive reactor graphite.
2015
148 pp
10.1016/j.jnucmat.2014.03.018
10.1007/s10512-005-0062-4
10.1016/j.commatsci.2013.03.026
10.1007/s10512-017-0263-7
Technology and installation for burning of irradiated reactor graphite.
VA
Kasheev
author
2013
text
Atomnaya Energiya
2013
122
4
210
213
10.1016/j.jnucmat.2015.01.063
VD
Nefedov
author
1960
Chemical modification due to induce by reaction (n, p).
1960
347 pp
10.1016/j.carbon.2016.04.024
Analysis of facility of potential hazard reduction of radioactive waste under thermal treatment. Izvestiya TPU.
AO
Pavliuk
author
2017
text
Inzhiniring georesursov
2017
328
8
24
32
VP
Rublevskij
author
2004
The role of carbon-14 in technogeneous irradiation of people.
2004
197 pp
MG
Sklyar
author
1984
Physical and chemical foundation of agglomeration.
1984
201 pp
10.1103/PhysRevLett.111.095501
The reactor graphite: development, production and properties.
YuS
Virgil’ev
author
2006
text
Rossijskij himicheskij zhurnal
2006
2006
1
4
12
10.1016/j.nucengdes.2013.09.007
About theory of reaction on porous or powdery material.
YaB
Zeldovich
author
1939
text
Zhurnal fizicheskoj himii
1939
13
2
163
168
10.1016/j.aop.2006.04.011
10.3897/nucet.4.29855
https://nucet.pensoft.net/article/29855/
https://nucet.pensoft.net/article/29855/download/pdf/
https://nucet.pensoft.net/article/29855/download/xml/
Issues associated with handling irradiated graphite of uranium-graphite nuclear reactors are examined. It is demonstrated that selection of approaches, methods and means for handling irradiated graphite are determined by the form of occurrence and binding energy of long-lived 14C radionuclide with graphite crystalline lattice. The purpose of the present study is the determination of possible chemical compounds in which 14C can be found and assessment of fastness of its binding in the structure of irradiated graphite. Indigent and foreign experience of handling graphite radioactive wastes was analyzed, calculations and measurements were performed. Information was provided on the channels of accumulation of 14C in the structure of reactor graphite and it was demonstrated that the largest quantities of the radionuclide in question are generated according to the reaction 14N(n, p)14C. Here, most part of radioactive carbon is generated on 14N nuclei found in the form of impurities in non-irradiated graphite and in the composition of gas used for purging nuclear reactor in the process of operation. 14C radionuclide generated according to 14N(n, p)14C nuclear reaction is localized in the near subsurface graphite layer (in the near subsurface layer of pores) at the depth of not more than 50 nm. Analysis was performed of possible chemical compounds which may incorporate radioactive carbon. It was established that the form of occurrence is determined by the operational properties of specific graphite element in the reactor core. 14C binding energy in the structure of irradiated graphite was evaluated and depth of its penetration in the structure was calculated. It was established that selective extraction of this radionuclide is possible only under elevated temperatures in weakly oxidizing environment which is explained by the binding energy reaching up to 800 kJ/mole in the process of chemical sorption of 14C on the surface of graphite and depth of its occurrence equal to ~ 70 nm in the course of ion implantation. It was demonstrated that radioactive carbon generated according to 13C(n, γ)14C nuclear reaction is uniformly distributed among graphite elements and possesses binding energy ~477 kJ/mole. Its selective extraction is possible only under the condition of destruction of graphite crystalline lattice and organization of the process of isotopic separation. The obtained results allow recommending the most efficient methods of handling irradiated graphite during decommissioning uranium-graphite reactors.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Uranium-graphite reactors
irradiated graphite
binding energy
binding strength
radionuclide
radioactive carbon
processing decontamination.
About chemical form and binding energy of 14C in irradiated graphite of uranium-graphite nuclear reactors
Research Article
10.3897/nucet.4.29858
2018-10-17
nucet
Obninsk Institute for Nuclear Power Engineering (INPE NRNU MEPhI), Obninsk, Russia
author
Yuferov, Anatoliy
2018-10-17
2018-10-17
2018
Nuclear Energy and Technology
2452-3038
4
1
57-63
2018
The TENDL Neutron Data Library and the 38-group neutron constant system TENDL038. VANT. Ser.
SN
Abramovich
author
2001
text
Yadernye konstanty
2001
1
11
26
AV
Alekseev
author
2012
2012
B
Beck
author
2006
2006
IN
Boboshin
author
1999
1999
IN
Boboshin
author
1994
1994
IN
Boboshin
author
1999a
1999a
MS
Drake
author
1970
1970
2009
2009
ENDF-6 (2009) Formats Manual, Data Formats and Procedures for the Evaluated Nuclear Data File ENDF/B-VI and ENDF/B-VII, BNL-90365-2009, edited by M.W Herman and A. Trkov, revised by M.W. Herman, A. Trkov and D.A. Brown (Dec. 2011).
Study on Relational ENDF Databases and Online Services.
TS
Fan
author
2005
text
Atomic Energy Science and Technology
2005
39
1
28
31
S
Holzner
author
2002
XSLT. The programmer’s library.
2002
544 pp
2012
2012
INDC(NDS)-0614 (2012) Further Development of EXFOR Summary Report of the Consultant’s Meeting. INDC International Nuclear Data Committee. IAEA Headquarters, 6–9 March, Vienna, Austria.
The format of the library of recommended data for the calculation of reactors. VANT. Ser.
VE
Kolesov
author
1972
text
Yadernye konstanty
1972
8
4
3
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Library of group constants BNAB_RF for reactor calculations and protection. Izvestiya vuzov.
VN
Koshcheev
author
2014
text
Yadernaya energetika
2014
3
83
101
Library of the estimated neutron data FOND-2.2. VANT. Ser.
VN
Koshcheev
author
2000
text
Yadernye konstanty
2000
2
40
70
NM
Larson
author
2007
2007
RE
MacFarlane
author
1996
1996
Priorities for the development of systems for the constant supply of reactor calculations and protection. Izvestiya vuzov.
GN
Manturov
author
2016
text
Yadernaya energetika
2016
2
133
142
GN
Manturov
author
2000
CONSYST code for neutron constants preparation. Scope statement: IPPE Preprint -2828.
2000
41 pp
10.1016/j.nds.2012.11.008
MN
Nikolaev
author
1984
Multigroup approximation in the theory of neutron transport.
1984
256 pp
K
Parker
author
1963
1963
VI
Plyaskin
author
2002
Reference-information interactive systems of nuclear physical data for various applications of nuclear physics.
2002
375 pp
Requirements for the Next Generation of Nuclear Databases and Services. J. Nucl. Sci. Technol., suppl.
V
Pronyaev
author
2001
text
2001
2
1476
1479
VV
Sinitsa
author
1993
1993
VV
Varlamov
author
2017
2017
VV
Varlamov
author
2001
2001
D
Woll
author
1968
1968
AG
Yuferov
author
2011
Converting the ROSFOND library to a relational database: IPPE Preprint-3194.
2011
28 pp
AG
Yuferov
author
2013a
Infologic model of the file of resonance parameters: IPPE Preprint -3233.
2013a
40 pp
AG
Yuferov
author
2013b
Relational model of the file of angular distributions: IPPE Preprint -3235.
2013b
20 pp
AG
Yuferov
author
2013c
Verification and validation of the file of resonant parameters in the relational format: IPPE Preprint -2828.
2013c
20 pp
VI
Zhuravlev
author
2009
2009
MN
Zizin
author
1974
Automation of reactor calculations.
1974
103 pp
10.3897/nucet.4.29858
https://nucet.pensoft.net/article/29858/
https://nucet.pensoft.net/article/29858/download/pdf/
https://nucet.pensoft.net/article/29858/download/xml/
The article considers the issues of converting the ENDF format systems of constants to relational databases. This conversion can become one of the tools facilitating the development and operation of factual information, techniques and algorithms in the field of nuclear data and, therefore, increasing the efficiency of the corresponding computational codes. The work briefly examines an infological model of ENDF libraries. The possible structure of tables of the corresponding relational database is described. The proposed database schema and the form of tables take into account the presence of both single and multiple properties of the isotopes under consideration. Consideration is given to the difference in organizational requirements for transferring constants from relational tables to programs and performing a visual analysis of data in tables by a physicist-evaluator. The conversion algorithms and results are described for the ROSFOND-A and ENDF/B-VII.1 libraries. It is shown that performing calculations directly in the DBMS environment has its advantages in terms of simplifying programming and eliminating the need to solve a number of problems on data verification and validation. Possible approaches are indicated to ensure operation of inherited software together with nuclear data libraries in the relational format. Some terminological refinements are proposed to facilitate constructing an infological model for ENDF format. The conversion programs and the ENDF/B-VII.1 library in the relational format are available on a public site.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
ENDF libraries
conversion
relational format.
Converting ENDF libraries into relational format
Research Article
10.3897/nucet.4.29859
2018-10-17
nucet
National Research Nuclear University MEPhI, Moscow, Russia
author
Uliyanin, Yury
National Research Nuclear University MEPhI, Moscow, Russia
author
Kharitonov, Vladimir
National Research Nuclear University MEPhI, Moscow, Russia
author
Yurshina, Daria
2018-10-17
2018-10-17
2018
Nuclear Energy and Technology
2452-3038
4
1
13-19
2018
EO
Adamov
author
2001
White Book of Nuclear Energy.
2001
270 pp
EO
Adamov
author
2007
Ecologically Pure Nuclear Energy.
2007
147 pp
da Rosa
Aldo Vieira
author
2005
2005
EN
Avrorin
author
2012
Strategy Concepts of Nuclear Energy Development in Russia at 21st Century.
2012
62 pp
VD
Borisevitch
author
2005
Basic Physics of Isotopes Separation in Gas Centrifuge.
2005
320 pp
2017
2017
BP Energy Outlook (2017) BP Energy Outlook. http://www.bp.com/content/dam/bp/pdf/energy-economics/energy-outlook-2017/bp-energy-outlook-2017.pdf
2017
2017
BP Statistical Review of World Energy (2017) BP Statistical Review of World Energy. http://www.bp.com/en/global/corporate/energy-economics/statistical-review-of-world-energy.html
2014
2014
Energy [R] Evolution (2014) A Sustainable USA Energy Outlook. Greenpeace International; European Renewable Energy Council (EREC); Global Wind Energy Council (GWEC). Report 3rd edition, USA energy scenario, 87 pр.
MK
Hubbert
author
1956
1956
2014
INPRO Manual. IAEA Nuclear Energy Series No. NG-T-4.4.
2014
103 pp
INPRO Methodology for Sustainability Assessment of Nuclear Energy Systems: Economics (2014) INPRO Manual. IAEA Nuclear Energy Series No. NG-T-4.4.IAEA, Vienna, 103 pp.
VV
Kharitonov
author
2014
Dynamic of nuclear energy development. Economic models.
2014
328 pp
Analytical Model of Metals Mining Dynamics.
VV
Kharitonov
author
2012
text
Tsvetnye Metally
2012
2012
10
20
24
VV
Kharitonov
author
2016
Long-term Exhausting Trends of Traditional Energy Resources and Nuclear Energy Perspectives.
2016
96 pp
Long-term and medium-term factors and scenarios of global energy system in the XXI century.
AE
Kontorovich
author
2014
text
Geologiya i Geofyzika
2014
55
5–6
689
700
NP
Laverov
author
2011
2011
V
Poplavsky
author
2011
2011
2017
Accelerating the global energy transformation.
2017
130 pp
REthinking Energy 2017 (2017) Accelerating the global energy transformation.International Renewable Energy Agency (IRENA), Abu Dhabi, 130 pp.
MN
Sinev
author
1987
Economics of Nuclear Energy. Fundamentals of Technology and Economics of Nuclear Fuel Production.
1987
320 pp
Peak Metals, Minerals, Energy, Wealth, Food and Population; Urgent Policy Considerations for A Sustainable Society.
HU
Sverdrup
author
2012
text
Journal of Environmental Science and Engineering B
2012
1
5
499
533
AV
Tarkhanov
author
2012
Modern Trends of World and Russian Uranium Industry (2007–2012). Minerals. Economic-geological series, No. 33. All-Russian Scientific-Research Institute of Mineral Raw Materials n.a. N. M.
2012
53 pp
The Generation IV International Forum. https://www.gen-4.org/gif/jcms/c_9260/public
2016
2016
Uranium 2016: Resources, Production and Demand (2016) A Joint Report by the Nuclear Energy Agency and the International Atomic Energy Agency, OECD, 550 pр.
EP
Velikhov
author
2010
Energy in Economics of the XXI century.
2010
176 pp
2017
2017
World Energy (2017) Issues Monitor 2017. World Energy Council, 156 pр.
2018
2018
World Nuclear Association. Information Library (2018) World Nuclear Association. Information Library. http://www.world-nuclear.org/information-library.aspx
VL
Zhivov
author
2012
Uranium : Geology, Extraction, Economics. All-Russian Scientific-Research Institute of Mineral Raw Materials n.a. N. M.
2012
304 pp
10.3897/nucet.4.29859
https://nucet.pensoft.net/article/29859/
https://nucet.pensoft.net/article/29859/download/pdf/
https://nucet.pensoft.net/article/29859/download/xml/
For the first time the analytical relationship was established between the nuclear energy generation worldwide and supply of NPPs with natural uranium, as conventional resources are expected to deplete by the end of this century. Forecast results include the dynamics of a potential increased shortage of conventional energy resources, such as hydrocarbon fuels (coal, oil, natural gas) and natural uranium, in the course of time due to a growing energy demand (at the rate of 1 to 2% per year), on the one hand, and the depletion of nonrenewable resources, on the other hand. The forecast is based on the current geological data on extractable hydrocarbon and uranium resources, and a mathematical model for the dynamics of nonrenewable resources production. The forecast shows that, with the present-day paradigm of handling the produced conventional energy sources, the reserves of these will be significantly depleted by the end of this century, and their production peaks are expected to be reached by the mid-century. In the event of state-of-art NPP designs, the dynamics of the installed capacity will follow the dynamics of the natural uranium depletion, and the NPP contribution to the supply of energy for the needs of humankind will go down while increasing at the same time the total shortage of conventional energy sources. By 2100, however, the contribution of nuclear power (based on thermal neutrons) to primary sources may reach 10%, since hydrocarbons will be depleted at a higher rate than uranium. Meanwhile, this amount of nuclear energy will be negligible, as compared to the demand for primary energy, after the 2040s even at the smallest possible rate of growth in demand (1%/year). A growing spread between the increasing energy demand and the decreasing supply of exhaustible conventional energy resources necessitates the evolution of nuclear fuel breeding (breeding of 239Pu from 238U and, possibly, 233U from 232Th) no later than the 2030s.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Nuclear energy
fuel burn-up
natural uranium
nonrenewable conventional energy resources
hydrocarbons
production dynamics
production rate
production peak
nuclear breeders.
Nuclear perspectives at exhausting trends of traditional energy resources
Research Article
10.3897/nucet.4.29837
2018-10-18
nucet
National Research Tomsk Polytechnic University, Tomsk, Russia
author
Shamanin, Igor
National Research Tomsk Polytechnic University, Tomsk, Russia
author
Bedenko, Sergey
https://orcid.org/0000-0003-4318-6338
National Research Tomsk Polytechnic University, Tomsk, Russia
author
Nesterov, Vladimir
National Research Tomsk Polytechnic University, Tomsk, Russia
author
Lutsik, Igor
National Research Tomsk Polytechnic University, Tomsk, Russia
author
Prets, Anatoly
2018-10-18
2018-10-18
2018
Nuclear Energy and Technology
2452-3038
4
1
79-85
2018
LP
Abagyan
author
1981
Group constants for the calculation of reactors and protection.
1981
139 pp
10.1097/00004032-198009000-00012
10.13182/NT69-A28386
Neutron radiation of 238PuO2 containing different amounts of 18O.
VA
Arkhipov
author
1972
text
Soviet Atomic Energy
1972
32
4
347
348
10.13182/NT73-A15883
10.13182/NSE71-18
D
Bell
author
1974
Theory of nuclear reactors.
1974
489 pp
Neutron yield of (α,n) reaction on oxygen.
VI
Bulanenko
author
1980
text
Soviet Atomic Energy
1980
47
1
531
534
Organization of the iterative process in the numerical reconstruction of the neutron spectrum in a multiplying system with a graphite retarder.
AV
Golovatskiy
author
2010
text
Russian Physics Journal
2010
53
11
10
14
Neutron Yield form the (α,n) Reaction in Be, B, C, O, F, Mg, Al, Si, and Granite Irradiation with Polonium a-particles.
VA
Gorshkov
author
1962
text
Soviet Atomic Energy
1962
13
123
135
The (α,n) cross section on 17O and 18O between 5 and 12,5 MeV.
LF
Hansen
author
1967
text
Nuclear Physics A
1967
98
1
23
32
10.1016/0168-9002(89)90579-2
10.13182/NT68-A26347
10.1080/00223131.2002.10875044
Neutron Radiation Field of the Irradiated Ceramic Nuclear Fuel of Different Types. Izvestiya vuzov.
IV
Shamanin
author
2010
text
Yadernaya Energetika
2010
2
97
103
10.13182/NSE71-A19667
10.13182/NT72-A16037
10.1007/s10512-015-9933-5
Group neutron cross sections for fission and radiation capture of transactinides. VANT. Ser.
AI
Voropayev
author
1979
text
Yadernye konstanty
1979
3
34
34
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Neutron yield for chemical compounds of actinides. Soviet Atomic Energy.
VA
Vukolov
author
1987
text
1987
62
4
271
276
D
West
author
1982
1982
WB
Wilson
author
2002
2002
10.1016/j.nimb.2010.02.091
10.3897/nucet.4.29837
https://nucet.pensoft.net/article/29837/
https://nucet.pensoft.net/article/29837/download/pdf/
https://nucet.pensoft.net/article/29837/download/xml/
An iteration method has been implemented to solve a neutron transport equation in a multigroup diffusion approximation. A thermoelectric generator containing plutonium dioxide, used as a source of thermal and electric power in spacecraft, was studied.
Neutron yield and multigroup diffusion approximation data was used to obtain a continuous and group distribution of neutron flux density spectra in a subcritical multiplying system.
Numerical multigroup approaches were employed using BNAB-78, a system of group constants, and other available evaluated nuclear data libraries (ROSFOND, BROND, BNAB, EXFOR and ENDSF).
The functions of neutron distribution in the zero iteration for the system of multigroup equations were obtained by approximating an extensive list of calculated and experimental data offered by the EXFOR and ENDSF nuclear data libraries. The required neutronic functionals were obtained by solving a neutron transport equation in a 28-group diffusion approximation. The calculated data was verified. The approach used is more efficient in terms of computational efforts (the values of the neutron flux density fractions converge in the third iteration). The implemented technique can be used in nuclear and radiation safety problems.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Subcritical system
neutron distribution function
neutron transport
multigroup diffusion approximation.
Solution of neutron-transport multigroup equations system in subcritical systems
Research Article
10.3897/nucet.4.30387
2018-10-18
nucet
JSC “SSC RF-IPPE”, Obninsk, Russia
author
Ivanov, Konstantin
JSC “SSC RF-IPPE”, Obninsk, Russia
author
Ivanov, Konstantin
JSC “SSC RF-IPPE”, Obninsk, Russia
author
Lavrova, Olga
JSC “SSC RF-IPPE”, Obninsk, Russia
author
Salaev, Sergey
JSC “SSC RF-IPPE”, Obninsk, Russia
author
Askhadullin, Radomir
2018-10-18
2018-10-18
2018
Nuclear Energy and Technology
2452-3038
4
1
73-78
2018
VV
Alekseev
author
2011
2011
RSh
Askhadullin
author
2003
2003
YuF
Balandin
author
1961
Structural Materials for the Facilities with Liquid Metal Coolants.
1961
207 pp
ME
Chernov
author
2003
2003
UR
Evans
author
1962
Corrosion and Oxidation of Metals. Transl. from Engl.
1962
885 pp
Oxidation Potential of Lead and Bismuth Melts. Izvesyiya vuzov.
BF
Gromov
author
1997
text
Yadernaya Energetika
1997
1997
6
14
18
KD
Ivanov
author
2005
2005
P
Kofstadt
author
1969
1969
O
Kubashevsky
author
1955
Oxidation of Metals and Alloys. Transl. from Engl.
1955
311 pp
IS
Kulikov
author
1986
Thermodynamics of Oxides. Handbook.
1986
342 pp
BL
Lipetsky
author
1985
Nonoxidation Heating of Rare Metals and Alloys in Vacuum.
1985
182 pp
10.1002/maco.201005871
OA
Molokanov
author
2002
2002
G
Mulier
author
2008
2008
2007
2007
OECD/NEA [Nuclear Science Committee] (2007) Handbook on Lead-Bismuth Eutectic Alloy and Lead Properties, Materials Compatibility, Thermal-Hydraulics and Technologies. OECD, 693 pp.
AE
Rusanov
author
2003
2003
NP
Zhuk
author
1976
The Course in the Theory of Corrosion and Protection of Metals.
1976
473 pp
10.3897/nucet.4.30387
https://nucet.pensoft.net/article/30387/
https://nucet.pensoft.net/article/30387/download/pdf/
https://nucet.pensoft.net/article/30387/download/xml/
The article considers the main methods for studying the process of oxidation of structural steels and evaluating their corrosion resistance in heavy liquid metal coolants (HLMC) under static and dynamic conditions. It is shown that the main disadvantage of these methods is the impossibility of evaluating the results in real time. The authors propose a new method for the experimental determination of the oxidation rate of steels in molten HLMCs, which makes it possible to measure the reaction rate without depressurizing the plant, by periodically injecting air doses of a known volume and monitoring its response time. As additional data in the research methodology, the results of a chemical-spectral analysis of the coolant and slags were provided after the experimental campaign completion as well as a metallographic analysis of steel samples to determine the oxide coating thickness and its comparison with the calculated value for the integral oxygen assimilation by the system. To implement the methodology, a laboratory facility was proposed, equipped with an oxygen thermodynamic activity (TDA) sensor of the RF-IPPE design. The sensor is certified by Gosstandart of Russia (certificate RU. 31.002 A No. 15464), registered in the State Register of Measuring Instruments (No. 25282-03) and approved for use in the Russian Federation.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Mass transfer
coolant
lead
lead-bismuth
structural steel
chemical-spectral analysis
oxidation of steels
oxide film
oxygen
oxygen activity sensor
oxidation kinetics.
Development of the technique for determination the rate of oxidation of structural steels in heavy liquid metal coolants
Research Article
10.3897/nucet.4.29844
2018-10-18
nucet
JSC “SSC RF-IPPE n.a. A.I. Leypunsky”, Obninsk, Russia
author
Morozov, Andrej
JSC “SSC RF-IPPE n.a. A.I. Leypunsky”, Obninsk, Russia
author
Pityk, Anna
JSC “SSC RF-IPPE n.a. A.I. Leypunsky”, Obninsk, Russia
author
Ragulin, Sergej
JSC “SSC RF-IPPE n.a. A.I. Leypunsky”, Obninsk, Russia
author
Sahipgareev, Azamat
JSC “SSC RF-IPPE n.a. A.I. Leypunsky”, Obninsk, Russia
author
Soshkina, Aleksandra
JSC “SSC RF-IPPE n.a. A.I. Leypunsky”, Obninsk, Russia
author
Shlyopkin, Aleksandr
2018-10-18
2018-10-18
2018
Nuclear Energy and Technology
2452-3038
4
1
65-71
2018
AS
Avanesyan
author
1980
Experimental study of the dynamic viscosity coefficient of boric acid aqueous solutions.
1980
20 pp
Thermal properties of boric acid aqueous solutions at 298–573 K.
ND
Azizov
author
1996
text
Teplofizika vysokih temperature
1996
34
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798
802
VM
Berkovich
author
2010
2010
Investigation of the thermal conductivity of electrolytes and porous materials saturated by a fluid aqueous solutions.
GG
Gusejnov
author
2007
text
Fizika, Baki, Elm
2007
13
1–2
13
25
Justification of HA- 2 passive reflooding systems design functions of advanced project NPP with VVER. Izvestiya vuzov.
SG
Kalyakin
author
2003
text
Yadernaya Energetika
2003
2003
2
94
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10.1007/s10512-014-9856-6
10.1115/ICONE17-75942
Design and experimental study of the non-condensable gases influence on the VVER steam generator model operation in condensing mode during beyond design basis accident. Izvestiya vuzov.
AA
Luk’yanov
author
2010
text
Yadernaya Energetika
2010
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4
172
182
Experimental justification for design functions of additional passive reflooding systems of the reactor VVER core.
AV
Morozov
author
2012
text
Teploenergetika
2012
2012
5
22
27
Experimental study of the steam generator VVER models in condensing mode.
AV
Morozov
author
2012a
text
Teploenergetika
2012a
2012
5
16
21
Analysis of the effect of operating factors on the operation of model of VVER steam generator in a mode of steam condensation. VANT. Ser.
AV
Morozov
author
2016
text
Nuclear and reactor constants
2016
2016
3
91
99
Experimental study of non-equilibrium thermal hydraulic processes in passive reflooding systems of reactor VVER core. Izvestiya vuzov.
OV
Remizov
author
2009
text
Yadernaya Energetika
2009
2009
4
115
123
II
Schmal
author
2015
2015
LS
Sterman
author
1982
Heat and Nuclear Power Plants. Tutorial for Universities.
1982
345 pp
10.1016/0029-5493(94)90111-2
2009
2009
WCAP-17021-NP (2009) Rev. 1. Summary of Tests to Determine the Physical Properties of Buffered and Un-buffered Boric Acid Solutions. https://www.nrc.gov/docs/ML1122/ML11220A169.pdf [accessed 13.01.2017]
10.1016/j.pnucene.2015.06.025
10.3897/nucet.4.29844
https://nucet.pensoft.net/article/29844/
https://nucet.pensoft.net/article/29844/download/pdf/
https://nucet.pensoft.net/article/29844/download/xml/
Process of boric acid mass transfer during accidents accompanied with rupture of circulation pipelines in VVER reactors of new generation equipped with passive safety systems are examined. Results of calculation of variation of boric acid concentration in VVER-TOI reactor in case of accident development process are presented. Positive effects of boric acid droplet entrainment on the processes of acid accumulation and crystallization in the reactor core are demonstrated. The obtained results allow formulating the conclusion on the possibility of these processes in the reactor core which may lead to the disruption of heat removal from fuel pins. Review of available published reference data on physical properties of boric acid solutions (density, viscosity, thermal conductivity) is given. It is established that available information is of too general nature and fails to cover the whole range of parameters (acid temperature, pressure and concentration) typical for potential emergency situation on NPP equipped with VVER reactor. Necessity of experimental study of processes of droplet entrainment under parameters typical for VVER emergency operation conditions, as well as investigation of thermal physics properties of boric acid within wide range of acid concentration values is required.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
VVER
emergency operation mode
boric acid
accumulation
droplet entrainment
thermal physics properties of boric acid solutions
density
viscosity
thermal conductivity.
Estimation influence of boric acid drop entrainment to its accumulation in the VVER reactor in the case of accident
Research Article
10.3897/nucet.4.30524
2018-11-26
nucet
Nizhny Novgorod State Technical University n.a. R.E. Alekseev, Nizhny Novgorod, Russia
author
Beznosov, Aleksander
Nizhny Novgorod State Technical University n.a. R.E. Alekseev, Nizhny Novgorod, Russia
author
Bokova, Tatyana
Nizhny Novgorod State Technical University n.a. R.E. Alekseev, Nizhny Novgorod, Russia
author
Bokov, Pavel
2018-11-26
2018-11-26
2018
Nuclear Energy and Technology
2452-3038
4
87-92
2018
AV
Beznosov
author
2012
2012
AV
Beznosov
author
2016
Technologies and key components for circuits of reactor plants and commercial and research facilities with lead and lead-bismuth coolants.
2016
488 pp
Experimental studies into the processes accompanying an intercircuit seal break in a steam generator with lead or lead-bismuth coolant and optimization of its design. Izvestiya vuzov.
AV
Beznosov
author
2006
text
Yadernaya energetika
2006
2006
4
3
11
AV
Beznosov
author
2005
2005
AV
Beznosov
author
2008b
2008b
AV
Beznosov
author
2002
2002
AV
Beznosov
author
2014b
2014b
An experimental study into the heat removal from HLMC by cooling fluid at atmospheric pressure. VANT. Ser.
AV
Beznosov
author
2016b
text
: Yaderno-reaktornye konstanty
2016b
2016
4
75
83
AV
Beznosov
author
2015
2015
Experimental studies into the processes accompanying an intercircuit seal break in a steam generator in heavy liquid metal cooled reactor plants. Izvestiya vuzov.
AV
Beznosov
author
2012a
text
Yadernaya energetika
2012a
2012
4
92
101
An experimental study into the regulation of the oxidizing potential in a lead and lead-bismuth coolant circuit using a gas mass exchanger. Izvestiya vuzov.
AV
Beznosov
author
2010
text
Yadernaya energetika
2010
2010
2
134
141
Tribotechnical studies of contact areas in the lead and lead-bismuth coolant environment.
AV
Beznosov
author
2012b
text
Vestnik mashinostroeniya
2012b
2012
1
43
46
AV
Beznosov
author
2007
2007
AV
Beznosov
author
2008
2008
10.1007/s10512-008-9011-3
Reactor plants with horizontal steam generators. VANT. Ser.
AV
Beznosov
author
2013
text
: Fizika yadernykh reaktorov
2013
2013
2
79
83
Experimental studies into natural circulation in lead and lead-bismuth cooled systems.
AV
Beznosov
author
2012b
text
Atomnaya energiya,
2012b
113
6
351
354
Experimental studies into the processes accompanying an intercircuit seal break in a steam generator within operationally safe limits in an HLMC reactor plant. Izvestiya vuzov.
AV
Beznosov
author
2016a
text
Yadernaya energetika
2016a
2016
2
154
162
AV
Beznosov
author
2014a
2014a
Ways to control the lead coolant oxygen thermodynamic activity. Proceedings of NNSTU n.a. R.Ye.
AV
Beznosov
author
2014a
text
Alekseyev
2014a
3
105
130
138
10.1115/ICONE21-15263
10.1115/ICONE21-15248
10.3897/nucet.4.30524
https://nucet.pensoft.net/article/30524/
https://nucet.pensoft.net/article/30524/download/pdf/
https://nucet.pensoft.net/article/30524/download/xml/
Small and medium sized lead and lead-bismuth cooled reactors currently under development in Russia are Generation IV reactors. This paper presents a review and new scientific and engineering solutions which are in line with the evolutionary development of small and medium sized reactor plants with heavy liquid metal coolants (HLMC).
A growing interest in small and medium sized reactor plants for transpolar applications, as well as for regional and other NPPs, and the emerging trend towards the substitution of coal-fired boiler stations for small modular reactors initiate R&D on new designs and operational solutions for fast neutron HLMC reactor plants. Such solutions are based on unique domestic experience of building and operating ground prototype test facilities and series lead-bismuth cooled reactor plants, as well as nuclear power units for various applications. These solutions provide for improved properties of advanced HLMC reactors, primarily in economic and safety terms, as compared to other small and medium sized reactor plants.
Theoretical and experimental work was undertaken at Nizhny Novgorod State Technical University (NNSTU) for justifying small and medium sized reactor plant designs with horizontal steam generators (BRS-GPG). Nonconventional scientific and engineering solutions have been considered aimed to improve the cost effectiveness and safety of HLMC NPP units, including for the localization of a potentially dangerous severe accident of the “intercircuit steam generator break” type. The review and integrated research results are presented which make it possible to justify nonconventional engineering solutions for the BRS-GPG reactor plant (reactor circuit circulation pattern, steam generator type, reactor circuit heat removal in standby and emergency modes, etc.).
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Small and medium sized reactor plant
key components
engineering solutions
heavy liquid metal coolants
intercircuit SG break
Components of small and medium sized HLMC reactor plant circuits
Research Article
10.3897/nucet.4.30525
2018-11-26
nucet
National Research Nuclear University MEPhI, Moscow, Russia
author
Kulikov, Yevgeny
National Research Nuclear University MEPhI, Moscow, Russia
author
Kulikov, Gennady
National Research Nuclear University MEPhI, Moscow, Russia
author
Apse, Vladimir
National Research Nuclear University MEPhI, Moscow, Russia
author
Shmelev, Anatoly
National Research Nuclear University MEPhI, Moscow, Russia
author
Geraskin, Nikolay
2018-11-26
2018-11-26
2018
Nuclear Energy and Technology
2452-3038
4
93-97
2018
2017
2017
Baratol (2017) Baratol. https://en.wikipedia.org/wiki/Baratol [accessed Apr. 11 2017]
Explosive Properties of Reactor-Grade Plutonium.
Mark J
Carson
author
1993
text
Science & Global Security
1993
4
11
128
2017
2017
Composition B (2017) Composition B. https://en.wikipedia.org/wiki/Composition_B [accessed Apr. 11 2017]
10.1016/0149-1970(82)90022-1
2016
2016
Dr Khan (2016) Dr Khan: Nuclear Smuggler Broke the Silence. http://www.atominfo.ru/news/air7423.htm [accessed Oct. 22 2016, in Russian]
Robert
Gillette
author
1977
1977
10.13182/NT80-A32527
1977
1977
International Conference on Nuclear Power and its Fuel Cycle (1977) Vol. 2. The nuclear fuel cycle, part 1. Salzburg, Austria, 2–13 May.
10.13182/NSE159-56
G
Kessler
author
2011
2011
10.13182/NSE07-A2644
G
Kessler
author
2008
2008
E
Kulikov
author
2009
2009
Calculational models for quantitative evaluation of proliferation protection for fissionable materials. Izvestiya vuzov.
YeG
Kulikov
author
2010
text
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2010
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2
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195
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Manelis
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BA
Nadykto
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Orlov
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Polymer-bonded explosive (2017) Polymer-bonded explosive. https://en.wikipedia.org/wiki/Polymer-bonded_explosive [accessed Apr. 11 2017]
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Stiller
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10.3897/nucet.4.30525
https://nucet.pensoft.net/article/30525/
https://nucet.pensoft.net/article/30525/download/pdf/
https://nucet.pensoft.net/article/30525/download/xml/
Since the closed nuclear fuel cycle suggests that plutonium is extracted from irradiated fuel and is recycled in nuclear reactors as part of the loaded fuel, proliferation resistance of fissile materials (plutonium) is becoming a problem of a practical significance. It is important to understand to what extent the physical and technical properties of fissile materials are capable to prevent these from being diverted to nonenergy uses. This paper considers the term ”proliferation resistance” from a physical and technical point of view with no measures taken for the physical protection, accounting and control of nuclear materials. Thus, proliferation resistance of plutonium means that it is technically impossible to fabricate a nuclear explosive device (NED) of the implosion type due to the overheating of the device’s components and the resultant NED failure.
The following conclusions have been made.
The assessment of the plutonium proliferation resistance is not justified where it relies on the analysis of an implosion-type NED excluding the use of modern heat-resistant and heat-conducting chemical explosives (CE) which are inaccessible.
Consideration of the asymptotic temperature profile in the NED components is not justified enough for the development of plutonium proliferation resistance recommendations.
No options enabling the slowdown of the NED warm-up process have been exhausted for analyzing the physical and technical factors that determine the proliferation resistance of plutonium.
General conclusion. The underlying rationale in a fundamental monograph by Dr. G. Kessler proved to be insufficiently valid, which has led to an unfounded inference as to the status of the plutonium proliferation resistance. The development of the procedures used and other factors taken into account are expected to increase the requirements to the content of the 238Pu isotope in plutonium for ensuring its proliferation resistance.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Plutonium
plutonium-238
proliferation resistance
nuclear explosive device
explosive
cryogenic temperatures
Computational model and physical and technical factors determining the plutonium proliferation resistance
Research Article
10.3897/nucet.4.30527
2018-11-26
nucet
Obninsk Institute for Nuclear Power Engineering (INPE NRNU MEPhI), Obninsk, Russia
author
Khorasanov, Georgiy
Obninsk Institute for Nuclear Power Engineering (INPE NRNU MEPhI), Obninsk, Russia
author
Samokhin, Dmitriy
Obninsk Institute for Nuclear Power Engineering (INPE NRNU MEPhI), Obninsk, Russia
author
Zevyakin, Aleksandr
JSC "SSC RF – IPPE", Obninsk, Russia
author
Zemskov, Yevgeniy
Nuclear Safety Institute of the Russian Academy of Sciences, Moscow, Russia
author
Blokhin, Anatoliy
2018-11-26
2018-11-26
2018
Nuclear Energy and Technology
2452-3038
4
99-102
2018
A
Aitkalieva
author
2017
2017
JF
Briesmeister
author
1997
1997
Selected macroscopic characteristics of medium sized fast reactors. Izvestiya vuzov.
GL
Khorasanov
author
2012
text
Yadernaya energetika
2012
2012
3
18
22
GL
Khorasanov
author
2017
2017
AV
Lopatkin
author
2013
Fuel cycle of large-scale nuclear power in Russia based on principles of fuel and radiation balance and nonproliferation. Dr. Tech. Sci. Diss.
2013
45 pp
A small fast lead cooled reactor for training purposes.
DS
Samokhin
author
2015
text
Izvestiya vuzov, Yadernaya energetika
2015
2015
3
135
141
Two plus one. A system of two components (VVER and BN) as the basis for the future and for solving the SNF problem.
VM
Troyanov
author
2016
text
Journal of Rosenergoatom
2016
2016
9
22
29
IV
Vaganov
author
2000
2000
10.3897/nucet.4.30527
https://nucet.pensoft.net/article/30527/
https://nucet.pensoft.net/article/30527/download/pdf/
https://nucet.pensoft.net/article/30527/download/xml/
The possibility for obtaining a hard neutron spectrum in small reactor cores has been considered. A harder spectrum than spectra in known fast sodium cooled and molten salt reactors has been obtained thanks to the selection of relatively small core dimensions and the use of metallic fuel and natural lead (natPb) coolant. The calculations for these compositions achieve an increased average neutron energy and a large fraction of hard neutrons in the spectrum (with energies greater than 0.8 MeV) caused by a minor inelastic interaction of neutrons with the fuel with no light chemical elements and with the coolant containing 52.3% of 208Pb, a low neutron-moderating isotope.
An interest in creating reactors with a hard neutron spectrum is explained by the fact that such reactors can be practically used as special burners of minor actinides (MA), and as isotope production and research reactors with new consumer properties. With uranium oxide fuel (UO2) substituted by metallic uranium-plutonium fuel (U-Pu-Zr), the reactors under consideration have the average energy of neutrons and the fraction of hard neutrons increasing from 0.554 to 0.724 MeV and from 18 to 28% respectively. At the same time, the one-group fission cross-section of 241Am increases from 0.359 to 0.536 barn, while the probability of the 241Am fission increases from 22 to 39%. It is proposed that power-grade plutonium resulting from regeneration of irradiated fuel from fast sodium cooled power reactors be used as part of the fuel for future burner reactors. It contains unburnt plutonium isotopes and some 1% of MAs which transmutate into fission products in the process of being reburnt in a harder spectrum. This will make it possible to reduce the MA content in the burner reactor spent fuel and to facilitate so the long-term storage conditions for high-level nuclear waste in dedicated devices.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Fast reactor
hard neutron spectrum
metallic uranium-plutonium fuel
natural lead coolant
americium-241
Lead reactor of small power with metallic fuel
Research Article
10.3897/nucet.4.30662
2018-11-26
nucet
JSC "SSC RF – IPPE named after A.I. Leypunsky", Obninsk, Russia
author
Zherdev, Gennady
JSC SSC RF – IPPE named after A.I. Leypunsky, Obninsk, Russia
author
Kislitsyna, Tamara
JSC SSC RF – IPPE named after A.I. Leypunsky, Obninsk, Russia
author
Nikolayev, Mark
2018-11-26
2018-11-26
2018
Nuclear Energy and Technology
2452-3038
4
103-109
2018
JP
Chaudat
author
1974
1974
10.13182/NSE70-A20720
AD
Frank-Kamenetskiy
author
1978
Simulation of neutron trajectories in Monte-Carlo calculations of nuclear reactors.
1978
96 pp
TS
Kislitsyna
author
2016
2016
TS
Kislitsyna
author
2016a
2016a
BNAB-RF library of group constants for nuclear reactor and shielding calculations.
VN
Koshcheyev
author
2014
text
Izvestiya vuzo, Yadernaya energetika
2014
3
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101
10.1016/j.nds.2010.11.001
BNAB-93 group constants system. Part 1. Nuclear constants for neutron and photon radiation field calculation. VANT. Ser.
GN
Manturov
author
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: Yadernye konstanty
1996
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Manturov
author
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1999
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MN
Nikolayev
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text
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Osetskaya
author
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Estimation of the fuel cost impact on the production program of energy companies in Russia.
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Osetskaya
author
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text
Fundamentalnye issledovaniya
2017a
2017
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381
386
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SV
Zabrodskaya
author
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: Yadernye konstanty,
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YeV
Zhemchugov
author
2017
2017
10.3897/nucet.4.30662
https://nucet.pensoft.net/article/30662/
https://nucet.pensoft.net/article/30662/download/pdf/
https://nucet.pensoft.net/article/30662/download/xml/
The ROCOCO system (Kislitsyna and Nikolayev 2016) is designed to supply constants for Monte-Carlo calculations of both neutron fields and the gamma fields they generate. The initial database of nuclear data used in the system is the Russian national library of evaluated neutron data (ROSFOND) (Nikolayev 2006, Zabrodskaya et al. 2007, RUSFOND 2017). ROCOCO is specific in that it enables the estimator to optimize the level of detail when describing neutron cross-section energy dependences. Cross-sections of key fuel, structural and coolant materials can be described in such detail as the evaluated data permits; cross-sections of secondary nuclides (minor actinides, fission products, etc.) can be described in a 299-group BNAB approximation (Manturov et al. 1996) with regard for the resonance self-shielding by subgroup method or without regard for self-shielding altogether. The energy dependence of gamma rays is described in a 127-group P5 approximation (Koshcheyev et al. 2014). Optimizing the level of detail makes it possible to reduce to a great extent the counting time with no major effect on the result and its error. Where desired, in the process of calculating the energy release in neutron reactions or in gamma-quanta formation matrices, contributions from the decay of radionuclides formed in these reactions (with a half-life of less than three years) can be taken into account. The energy dependence of the elastic scattering anisotropy is described in detail, or in the event of a group or subgroup description of cross-sections, by defining 33 boundaries of 32 equiprobable cosine intervals of the scattering angle. The thermalization effects in calculations of neutron fields are taken into account either in an ideal gas approximation or using 72-group thermalization matrices built based on thermalization files contained (if any) in the ROSFOND library.
It should be noted that the system contains descriptions of detailed dependences of elastic scattering cross-sections and angular distributions on all multi-isotope elements; the relationship between the scattering angle and the energy loss in this case is determined with the use of the energy-dependent effective atomic weight.
The system’s programs are written in the FORTRAN language. The system is easily integrated in Monte-Carlo codes.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Nuclear data
radiation field Monte-Carlo calculation
combination of detailed and group descriptions of cross-section energy dependence
ROCOCO: A constants supply system for Monte-Carlo reactor calculation
Research Article
10.3897/nucet.4.30663
2018-11-26
nucet
National Research Nuclear University MEPhI, Obninsk, Russia
author
Kazansky, Yury
Experimental Scientific Research and Methodology Center “Simulation Systems” (SSL), Obninsk, Russia
author
Slekenichs, Yanis
2018-11-26
2018-11-26
2018
Nuclear Energy and Technology
2452-3038
4
111-118
2018
AM
Afrov
author
2006
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488 pp
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Cherkashov
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631 pp
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2004
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FGUP “Rossiyskiy gosudarstvennyy kontsern po proizvodstvu elektricheskoy i teplovoy energii na atomnykh stantsiyakh” (2004) Guidance document. Methods for calculating neutron physics characteristics from the data of physical experiments on power units of nuclear power plants with VVER-1000 reactors. RD EO 0151-2004. Moscow, 101 pp. [in Russian]
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Kazansky
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Khetrik
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Kirillov
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2007
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10.3897/nucet.4.30663
https://nucet.pensoft.net/article/30663/
https://nucet.pensoft.net/article/30663/download/pdf/
https://nucet.pensoft.net/article/30663/download/xml/
It is assumed by the authors of the present paper that with growing contribution of nuclear power in the production of electricity, nuclear power plants will be used to a higher degree in a manoeuvrable mode of operation rather than in the base-load mode. In other words, change of power from the nominal level to that of coverage of auxiliary loads will be becoming quite common and not so rare event as scheduled reactor shutdowns for fuel reloading or preventive works. There exist well-known problems in the use of nuclear reactors in the manoeuvrable operation mode, which include the task shared by all types of nuclear reactors. It is advisable to have a unified indicator weakly power-dependent and fairly easy to measure, which would make it possible to formulate the judgement about the nature of the transient processes within the entire power range and to assess the reactivity required for changing the power level by the preset value. Power reactivity coefficient (PRC) can be used as such indicator. Analysis was made of existing definitions and understanding of PRC in relevant references. It turned out that there is no generally accepted definition of the PRC. Based on the performed study, the following definition was suggested: the PRC is the ratio of the low reactivity introduced into the reactor to the power increment at the end of the transient process. It is assumed here that variation of reactivity is dependent on the energy released in nuclear fission but is not related to the changes of reactivity induced by feedback signals in the automatic reactor power control system.
Analysis of the relationship between the PRC and temperature coefficients and technological parameters associated with the steady-state control program was performed taking the above suggested definition into account. PRC calculations were performed using the simplest model of VVER-1000 type power reactor. It was found that PRC is weakly power-dependent.
The purpose of the present study is to investigate dependence of PRC on the temperature reactivity effects and on the technological parameters associated with the steady-state control program of the power unit, using the example of VVER-1000. Effects of PRC on the static and dynamic power reactor operation modes are analyzed.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Nuclear power plants
power reactivity coefficient
temperature reactivity coefficients
nuclear reactor dynamics
Power coefficient of reactivity: definition, interconnection with other coefficients of reactivity, evaluation of results of transients in power nuclear reactors
Research Article
10.3897/nucet.4.30677
2018-11-26
nucet
Kola Science Centre of the Russian Academy of Sciences, Apatity, Russia
author
Naumov, Vadim
Kola Science Centre of the Russian Academy of Sciences, Apatity, Russia
author
Gusak, Sergey
Kola Science Centre of the Russian Academy of Sciences, Apatity, Russia
author
Naumov, Andrey
2018-11-26
2018-11-26
2018
Nuclear Energy and Technology
2452-3038
4
119-125
2018
EO
Adamov
author
2015
2015
PN
Alekseev
author
2016
2016
SV
Egorov
author
2016
2016
TR
England
author
1994
1994
2012
2012
IAEA-TECDOC-1536 (2012) Status of Small Reactor Designs without On-site Refueling. http://www-pub.iaea.org/MTCD/Publications/PDF/te_1536_web.pdf [accessed 18 Nov. 2012]
SV
Ignatiev
author
2007
2007
NN
Klimov
author
2013
2013
Design solutions of RITM-200 reactor installation intended for ensuring environmentally safe and cost-effective operation of multipurpose nuclear icebreaker on Arctic routes.
KYu
Knyazevsky
author
2014
text
Arktika: ekologiya i ekonomika
2014
3
15
86
91
RA
Konyukhov
author
2015
2015
The experience of construction and operation of reactor plants for civilian ships.
VI
Makarov
author
2000
text
Atomnaya energiya
2000
89
3
79
188
Reactor units for power supply to remote and isolated regions: selection problem.
NN
Melnikov
author
2015
text
Vestnik MGTU
2015
18
2
198
208
VA
Naumov
author
1996
The software package KRATER for calculation of neutronics of thermal nuclear reactors: Preprint IPE-14. Minsk.
1996
39 pp
VV
Petrunin
author
2015
2015
OB
Samoylov
author
2005
2005
BG
Saneev
author
2011
2011
2015
2015
Strategy of development of Arctic zone (2015) Strategy of development of Arctic zone of the Russian Federation and ensuring national security for the period until 2020. http://docs.cntd.ru/document/499002465 [assessed 5 Jan. 2015, in Russian]
Development of fuel rods for reactor cores of floating power plants (FPP) and low-power nuclear power plants (LPNPP): state and potential. VANT. Ser.
AV
Vatulin
author
2005
text
Materialovedenie i novye materialy
2005
2
65
146
148
Calculation of SVBR-100 reactor fuel load lifetime taking into account control and burn-up compensation rod displacement. VANT. Ser.
AV
Voronkov
author
2009
text
Obespechenie bezopasnosti AES
2009
24
38
43
NI
Voropay
author
2015
2015
10.3897/nucet.4.30677
https://nucet.pensoft.net/article/30677/
https://nucet.pensoft.net/article/30677/download/pdf/
https://nucet.pensoft.net/article/30677/download/xml/
The purpose of the present study is the investigation of mass composition of long-lived radionuclides accumulated in the fuel cycle of small nuclear power plants (SNPP) as well as long-lived radioactivity of spent fuel of such reactors. Analysis was performed of the published data on the projects of SNPP with pressurized water-cooled reactors (LWR) and reactors cooled with Pb-Bi eutectics (SVBR). Information was obtained on the parameters of fuel cycle, design and materials of reactor cores, thermodynamic characteristics of coolants of the primary cooling circuit for reactor facilities of different types. Mathematical models of fuel cycles of the cores of reactors of ABV, KLT-40S, RITM-200M, UNITERM, SVBR-10 and SVBR-100 types were developed. The KRATER software was applied for mathematical modeling of the fuel cycles where spatial-energy distribution of neutron flux density is determined within multi-group diffusion approximation and heterogeneity of reactor cores is taken into account using albedo method within the reactor cell model. Calculation studies of kinetics of burnup of isotopes in the initial fuel load (235U, 238U) and accumulation of long-lived fission products (85Kr, 90Sr, 137Cs, 151Sm) and actinoids (238,239,240,241,242Pu, 236U, 237Np, 241Am, 244Cm) in the cores of the examined SNPP reactor facilities were performed. The obtained information allowed estimating radiation characteristics of irradiated nuclear fuel and implementing comparison of long-lived radioactivity of spent reactor fuel of the SNPPs under study and of their prototypes (nuclear propulsion reactors). The comparison performed allowed formulating the conclusion on the possibility in principle (from the viewpoint of radiation safety) of application of SNF handling technology used in prototype reactors in the transportation and technological process layouts of handling SNF of SNPP reactors.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Arctic regions of Russia
small nuclear power plants
reactors
spent nuclear fuel
fuel cycle
radioactivity
Small nuclear power plants for power supply in arctic regions: assessment of spent nuclear fuel radioactivity
Research Article
10.3897/nucet.4.30771
2018-11-26
nucet
JSC “Pilot and Demonstration Center for Decommissioning of Uranium-Graphite Nuclear Reactors”, Seversk, Russia
author
Pavliuk, Alexander
https://orcid.org/0000-0001-8897-6910
JSC “Pilot and Demonstration Center for Decommissioning of Uranium-Graphite Nuclear Reactors”, Seversk, Russia
author
Kotlyarevskiy, Sergey
JSC “Pilot and Demonstration Center for Decommissioning of Uranium-Graphite Nuclear Reactors”, Seversk, Russia
author
Bespala, Evgeny
National Research Tomsk Polytechnic University, Tomsk, Russia
author
Bespala, Yuliya
2018-11-26
2018-11-26
2018
Nuclear Energy and Technology
2452-3038
4
127-133
2018
10.1016/j.nucengdes.2013.09.007
10.1016/j.jnucmat.2014.03.018
2006
2006
EPRI (2006) Graphite Decommissioning: Options for Graphite Treatment, Recycling, or Disposal, including a discussion of Safety-Related Issues. Technical Report 1013091. https://pdfs.semanticscholar.org/1367/38dccadbc420b7a112af9dd4c3b6885c6e5d.pdf
10.1016/j.nucengdes.2008.02.010
2007
2007
IAEA (2007) Disposal aspects of low and intermediate level decommissioning waste. Technical Report IAEA-TECDOC-1572. Vienna, IAEA Publ. http://www-pub.iaea.org/MTCD/publications/PDF/TE_1572_companion_CD_web.pdf [accessed Sept. 4 2017]
10.1016/j.carbon.2013.02.053
10.1016/j.jnucmat.2015.01.063
10.1016/j.net.2015.08.007
10.1016/j.carbon.2016.04.024
10.1016/j.jenvrad.2018.01.005
10.1007/s10512-014-9826-z
AA
Romenkov
author
2011
2011
VP
Rublevsky
author
2004
2004
10.1016/0008-6223(74)90008-6
10.1016/j.nimb.2014.02.040
10.5516/NET.06.2012.025
Treatment and disposal of irradiated graphite and other carbonaceous waste.
W
Von Lensa
author
2011
text
Atw-International Journal for Nuclear Power
2011
57
263
269
10.1016/j.jnucmat.2013.02.027
10.1016/j.jenvrad.2017.01.022
10.3897/nucet.4.30771
https://nucet.pensoft.net/article/30771/
https://nucet.pensoft.net/article/30771/download/pdf/
https://nucet.pensoft.net/article/30771/download/xml/
Aspects of handling irradiated graphite during decommissioning uranium-graphite reactors (UGR) of different types were investigated. It was demonstrated that handling reactor graphite is complicated by the presence in the composition of graphite of long-lived radionuclides, especially 14C, which may get entrained in biological cycles since carbon constitutes one of the main components of biological chains. Practical implementation of the process of selective separation of 14С can significantly reduce potential danger represented by graphite radioactive wastes due to the reduction of graphite activity as related to the isotope in question, as well as due to the reduction of the leaching rate by separating 14С isotope which is the most weakly bound within the graphite structure. Conclusion was formulated that analytical measurement methodologies and calculation methods allow reliably estimating only the total quantity of 14C accumulated in graphite, the contribution of 14C accumulation channel from 13C(n, γ)14C reaction, as well as the total contribution of 14N(n, p)14C reaction on nitrogen impurities and on nitrogen contained in purge gas. Method was suggested for estimating the values of contributions of different channels of accumulation on nitrogen impurities and nitrogen contained in purge gas using IRT-T research reactor (Tomsk, Tomsk Region). Parallel irradiation of batches of samples of non-irradiated (fresh) reactor-grade graphite contained in different gaseous media constitutes the basis of the study. Algorithm was suggested for calculating contributions of all channels of 14C accumulation according to the results of measurements to be obtained in the proposed studies. Recommendations were formulated on the use of all brands of graphite applied for manufacturing elements of graphite stacks of uranium-graphite reactors designed in Russia for determining selectively separated fraction of 14C for all types of graphite radioactive wastes by the companies in the RF which operated (are operating) the uranium-graphite reactors. Time of exposure of samples of irradiated graphite in the GEK-4 horizontal experimental channel of the IRT-T reactor was calculated and was found to be equal to ~ 10 days. Methodology was suggested for conducting a series of experiments for determining the values of contributions of 14C accumulation channels in the irradiated reactor graphite. The methodology suggested can be applied for determining fraction of selectively separated 14C in irradiated graphite elements of practically all uranium-graphite nuclear reactors, including reactors operated abroad Russia, under the condition of maintaining carbon dioxide gas atmosphere in one of the irradiated containers.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Irradiated reactor-grade graphite
uranium-graphite reactor
research reactor
IRT-T
Potential of application of IRT-T research reactor as the solution of the problem of graphite radwaste disposal
Research Article
10.3897/nucet.4.30774
2018-11-26
nucet
State Scientific Center – Institute for Physics & Power Engineering (SSC RF-IPPE), Obninsk, Russia
author
Mukhamadeev, Ruben
State Scientific Center – Institute for Physics & Power Engineering (SSC RF-IPPE), Obninsk, Russia
author
Parafilo, Leonid
State Scientific Center – Institute for Physics & Power Engineering (SSC RF-IPPE), Obninsk, Russia
author
Baranaev, Yury
State Scientific Center – Institute for Physics & Power Engineering (SSC RF-IPPE), Obninsk, Russia
author
Suvorov, Albert
2018-11-26
2018-11-26
2018
Nuclear Energy and Technology
2452-3038
4
135-142
2018
YuD
Baranaev
author
1993
1993
10.1007/BF00844625
2016
In-depth Safety Analysis Report. JSC Concern Rosenergoatom, JSC SSC IPPE, JSC Atomenergoproekt, JSC Izhorskiye zavody.
2016
4971 pp
Bilibino NPP. Unit 4. (2016) In-depth Safety Analysis Report. JSC Concern Rosenergoatom, JSC SSC IPPE, JSC Atomenergoproekt, JSC Izhorskiye zavody.Moscow, Rosenergoatom, 4971 pp. [in Russian]
10.13182/NT83-1
JR
Deen
author
1995
WIMS-D/4 User Manual. Rev.0.
1995
95 pp
2009
IAEA SAFETY STANDARDS SERIES No.
2009
84 pp
Deterministic Safety Analysis for Nuclear Power Plants (2009) IAEA SAFETY STANDARDS SERIES No.SSG-2, STI/PUB/1428, IAEA, Vienna, 84 pp.
Dolgov VV, Ilyin YuV, Ryabov VV (1995, 1996) Results of the Bilibino NPP fuel element testing in conditions simulating an accident with complete loss of coolant and failure of all active cooldown channels/The 4th Interdepartmental Conference on Reactor Material Science, 15-19 May 1995, Dimitrovgrad. Collection of Reports in 2 Volumes. Dimitrovgrad. NIIAR Publ., V. 1. Fuel and Fuel Elements of Power Reactors, 165–177. [in Russian]
Fletcher CD, Schultz RR RELAP5/MOD3 Code Manual. NUREG/CR-5535, INEL-95/0174, v. 2, 293 pp.
2003
2003
IAEA-TECDOC-1345 (2003) Fuel Failure in Water Reactors: Causes and Mitigation (Proceedings of Technical Meeting, Bratislava, 2002), IAEA, Vienna, 165 pp.
2005
Structural Behavior of Fuel Assemblies for Water Cooled Reactors (Proc. Tech. Mtg.
2005
324 pp
IAEA-TECDOC-1454 (2005) Structural Behavior of Fuel Assemblies for Water Cooled Reactors (Proc. Tech. Mtg.Cadarache, France, 2004), IAEA, Vienna, 324 pp.
2010
2010
NF-T-2.1, STI/PUB/1445 (2010) Review of fuel failures in water cooled reactors. IAEA, 178 pp.
2015
2015
NP-001-2015 (2015) General Provisions for Ensuring NPP Safety. Moscow. Rostekhnadzor Publ. [in Russian]
2007
2007
NP-082-07 (2007) Nuclear Safety Regulations for Reactor Facilities of Nuclear Power Plants. Moscow. Rostekhnadzor Publ., 26 pp. [in Russian]
2005
2005
RB-001-05 (2005) Recommendations on the Content of the In-depth Safety Analysis Report for Operated Nuclear Power Units (OUOB AS). Rostekhnadzor Publ, Moscow. [in Russian]
2012
Certification Passport as Applied to Calculations for the EGP-6 Reactor. No. 317, dated 9.10.2012.
2012
6 pp
RELAP5/MOD3.2 (2012) Certification Passport as Applied to Calculations for the EGP-6 Reactor. No. 317, dated 9.10.2012.Rostekhnadzor Publ., Moscow, 6 pp. [in Russian]
AG
Samoylov
author
1982
Dispersion Fuel Elements. Vol. 1. Materials and Technologies.
1982
224 pp
AG
Samoylov
author
1982a
1982a
AG
Samoylov
author
1996
Fuel Elements of Nuclear Reactors.
1996
400 pp
VN
Sharapov
author
1997
Neutronic fundamentals of water-graphite reactors with tubular fuel elements. Dr. Sci. thesis, specialization 05.14.03.
1997
75 pp
Suvorov AP, Baranaev YuD, Moseev LI, Kozmenkov YaK (1996, 1998) Experimental and Calculation Investigation of Fission Product Release from Fuel of Water Cooled Reactors on Initial Stage of Severe Accident. Report on TCM Design measures for prevention and mitigation of severe accident at advanced water-cooled reactors, 21-25 October, Vienna. IAEA-TECDOC-1020, IAEA, Vienna, 27–37.
1998
Review of Fuel Failures in Water Cooled Reactors.
1998
167 pp
Technical Reports Series No. 388, STI/DOC/010/388 (1998) Review of Fuel Failures in Water Cooled Reactors.IAEA, Vienna, 167 pp.
10.3897/nucet.4.30774
https://nucet.pensoft.net/article/30774/
https://nucet.pensoft.net/article/30774/download/pdf/
https://nucet.pensoft.net/article/30774/download/xml/
Analysis was performed of dynamic phase of severe accident of the EGP-6 reactor of the Bilibino NPP, due to uncontrolled reactivity insertion initiated by withdrawal of two pare of automatic control rods with followed by full failure of reactor emergency protection system. This initial event leads to promt increasing of reactor core power up to 450% of nominal value with short period, coupled with rise of temperature of fuel, pressure and temperature of coolant. These factors lead to crisis of heat exchange with subsequent ruptures tubes of fuel assemblies and coolant blow down into graphite stack. All its lead to rise of pressure in reactor shell and damage of it, outflow of steam-water mixture through up-reactor area to ventilation system, communication corridors and reactor hall and further – to atmospheric release. Transient processes were calculated using code RELAP5/Mod3.2. It was considered stages of processes of fuel damage and evaluated dynamic of a number and degree of damaged fuel assembles. They were grouped on burn-up and for each group it was performed analysis of dynamic of damage values. Further it was considered processes of yield of fission products from damaged fuel with models, based on experimental data on yield of fission products from fuel material, used in assembles of Bilibino NPP fuel type (fuel tubes with steel cladding, where fuel material is grits of uranium dioxide in magnesium), under condition of severe accident, especially performed in SSC IPPE. Transport of fission products with steam and air up to release points was evaluated with models, based on experimental data of fission product transport through graphite stack under conditions of severe accident, also especially performed in SSC IPPE. Evaluation of source term was performed in accordance with accident dynamic and assumed modes of release for conservative and most possible approaches. It was noted good self-protection property of EGP-6 reactor under severe beyond design basis accident condition.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Severe beyond design basis accident
heat exchange crisis
fuel damage dynamic
fission products yields
radioactive source term
Analysis of a severe beyond design basis accident for the EGP-6 reactor of the Bilibino NPP. Radioactive source term determination
Research Article
10.3897/nucet.4.30775
2018-11-26
nucet
Obninsk Institute for Nuclear Power Engineering (INPE NRNU MEPhI), Obninsk, Russia
author
Leskin, Sergey
Joint Stock Company «Scientific Technical Center «Diaprom», Moscow, Russia
author
Shvetsov, Dmitry
Joint Stock Company «Scientific Technical Center «Diaprom», Moscow, Russia
author
Trykov, Evgeny
Obninsk Institute for Nuclear Power Engineering (INPE NRNU MEPhI), Obninsk, Russia
author
Puzakov, Aleksey
2018-11-26
2018-11-26
2018
Nuclear Energy and Technology
2452-3038
4
141-147
2018
System for monitoring and diagnostics of NPP unit operational conditions.
AA
Abagyan
author
1987
text
Atomnaya energiya
1987
63
311
315
GV
Arkadov
author
2010
Systems diagnosis of VVER.
2010
391 pp
10.1016/0005-1098(88)90073-8
10.1016/0005-1098(90)90018-D
K
Fukunaga
author
1972
Introduction to Statistical Pattern Recognition.
1972
375 pp
A review of on-line diagnostic aids for nuclear power plant operators. Nucl.
MR
Herbert
author
1984
text
Energy
1984
23
4
259
264
10.1016/0005-1098(84)90098-0
Algorithm development for detecting abnormality of NPP equipment conditions based on technological testing results. Izvestiya vuzov.
ST
Leskin
author
1997
text
Yadernaya energetika
1997
1997
4
4
12
Analysis of VVER-1000 main circulation pump conditions in operation. Isvestya vuzov.
ST
Leskin
author
2016
text
Yadernaya energetika
2016
2016
4
12
22
Analysis of safety system pumps conditions based on their testing results. Isvestya vuzov.
ST
Leskin
author
2017
text
Yadernaya energetika
2017
2017
1
42
50
R
Patton
author
1989
Fault Diagnosis in Dynamic Systems – Theory and Applications.
1989
360 pp
Review of applications of expert methodology systems for nuclear power engineering.
VI
Pavelko
author
1990
text
Atomnaya tekhnika za rubezhom
1990
1990
11
1
8
Evaluating operator support system in realistic conditions at hammlab.
C
Reisen
author
1988
text
Nuclear Engineering International,
1988
33
402
39
41
V
van Ryzin
author
1980
Classification and Clustering.
1980
365 pp
A new criterion for optimal classification.
Gu
Tao
author
1982
text
Pattern Recognition
1982
2
1063
1065
J
Tu
author
1978
Pattern Recognition Principles.
1978
412 pp
A probabilistic analysis method of evaluate the effect of human factors on plant safety. Nucl. Tehnol.
Hiroshi
Ujita
author
1986
text
1986
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Robert E
Urig
author
1991
1991
S
Weiss
author
1988
1988
10.1016/0005-1098(76)90041-8
10.3897/nucet.4.30775
https://nucet.pensoft.net/article/30775/
https://nucet.pensoft.net/article/30775/download/pdf/
https://nucet.pensoft.net/article/30775/download/xml/
Acoustic leak control systems (for instance, SAKT) are used at present for controlling leak tightness of equipment and pipelines, as well as for detecting in timely manner coolant leaks from the primary cooling circuit of nuclear reactor installations (NRI) during operation of power unit on different power levels in the modes of normal operation and during disturbances of normal operation. Time averaged dispersion of acoustic signal is used as the main diagnostic indicator for detecting leaks in these systems. Sensitivity of this indicator is determined by the exceedance by the signal of the preset threshold value which is defined in accordance with the background. Here, background values of acoustic signal depend on the operational modes of the equipment and do not allow in many cases determining coolant leak during early stages of leak development.
New approach to the formation of diagnostic indicators for detecting loss of sealing in the circuit during early stage of development of coolant leak is suggested.
Methodology for obtaining diagnostic indicators is based on the processing in different frequency bands of acoustic signal accompanying coolant leakage from the pipeline using the method of principal components.
Efficiency of the developed methodology of coolant leak detection is illustrated by processing acoustic signals for experimental facility modeling coolant leakage in case of loss of sealing of the circuit.
Even in the presence of significant acoustic background sensitivity of the method allows detecting leaks with significantly lower flow rates (up to five times smaller) than the conventional processing of acoustic signals.
Implementation of the developed methodology will not require significant expenditures for upgrading already existing leak control systems operated at present on different NPPs.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
control of leak tightness of equipment
dispersion of acoustic signal
modeling coolant leaks
method of principal components
additional diagnostic indicators
pattern identification
Analysis of acoustic leak signals for enhancing sensitivity of control due to the creation of effective diagnostic indicators
Research Article
10.3897/nucet.4.30777
2018-11-26
nucet
JSC “Afrikantov OKBM”, Nizhny Novgorod, Russia
author
Kulikov, Aleksey
JSC “Afrikantov OKBM”, Nizhny Novgorod, Russia
author
Lepyokhin, Andrey
JSC “Afrikantov OKBM”, Nizhny Novgorod, Russia
author
Polunichev, Vitaly
2018-11-26
2018-11-26
2018
Nuclear Energy and Technology
2452-3038
4
149-154
2018
VS
Chirkin
author
1968
Thermоphysical properties of materials of nuclear engineering.
1968
484 pp
VM
Belyaev
author
2016
2016
IE
Idelchik
author
1990
The reference book on hydraulic resistances.
1990
672 pp
2005
2005
Innovative Small and Medium Sized Reactors: Design Features, Safety Approaches and R&D Trends (2005) . Final report of a technical meeting held in Vienna, 7–11 June 2004. IAEA-TECDOC-1451. IAEA, Vienna.
VS
Kuul
author
1992
1992
AN
Lepyokhin
author
2017
2017
AN
Lepyokhin
author
2010
2010
YuA
Migrov
author
2001
2001
1995
1995
RELAP5/mod3 (1995) . Сode Manual. Idaho National Engineering Laboratory, June. NUREG/CR 5535 V1–V5.
SL
Rivkin
author
1980
Thermоphysical properties of water and water steam.
1980
424 pp
2006
2006
Thermophysical properties database of material for Light Water Reactors and Heavy Water Reactors (2006) . Final report of a coordinated research project 1999–2005. IAEA-TECDOC-1496. IAEA, Vienna.
MV
Vorobyova
author
2016
2016
MV
Vorobyova
author
2008
2008
New generation reactor plant RITM-200 for the perspective nuclear icebreaker.
DL
Zverev
author
2012
text
Atomnaya energiya [Atomic Energy]
2012
113
6
323
328
10.3897/nucet.4.30777
https://nucet.pensoft.net/article/30777/
https://nucet.pensoft.net/article/30777/download/pdf/
https://nucet.pensoft.net/article/30777/download/xml/
The purpose of the work was to optimize the parameters of the spillage system equipped with a gas pressure hydroaccumulator for a ship pressurized water reactor in a loss-of-coolant accident. The water-gas ratio in the hydroaccumulator and the hydraulic resistance of the path between the hydroaccumulator and the reactor were optimized at the designed hydroaccumulator geometric volume.
The main dynamic processes were described using a mathematical model and a computational analysis. A series of numerical calculations were realized to simulate the behavior dynamics of the coolant level in the reactor during the accident – by varying the optimized parameters. Estimates of the minimum and maximum values of the coolant level were obtained: depending on the initial water-gas ratio in the hydroaccumulator at different diameters of the flow restrictor on the path between the hydroaccumulator and the reactor. These results were obtained subject to the restrictive conditions that, during spillage, the coolant level should remain above the core and below the blowdown nozzle. The first condition implies that the core is in safe state, the second excludes the coolant water blowdown. The optimization goal was to achieve the maximum time interval in which these conditions would be satisfied simultaneously.
The authors propose methods for selecting the optimal spillage system parameters; these methods provide the maximum time for the core to be in a safe state during a loss-of-coolant accident at the designed hydroaccumulator volume. Using these methods, it is also possible to make assessments from the early stages of designing reactor plants.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Accident
coolant leak
coolant level
core
safe state
optimization
spillage system
reactor
hydroaccumulator
design analysis
Analysis of mass transfer processes in a reactor during a loss-of-coolant accident
Research Article
10.3897/nucet.4.30779
2018-12-07
nucet
JSC RPA “CNIITMASH”, Moscow, Russia
author
Anosov, Nikolay
JSC RPA “CNIITMASH”, Moscow, Russia
author
Skorobogatykh, Vladimir
JSC RPA “CNIITMASH”, Moscow, Russia
author
Gordyuk, Lyubov
JSC RPA “CNIITMASH”, Moscow, Russia
author
Mikheev, Vasiliy
JSC RPA “CNIITMASH”, Moscow, Russia
author
Pogorelov, Yegor
JSC “SSC RIAR”, Dimitrovgrad, Russia
author
Shamardin, Valentin
2018-12-07
2018-12-07
2018
Nuclear Energy and Technology
2452-3038
4
155-161
2018
Express method for evaluating radiation resistance of weld joints as a function of chemical composition.
NP
Anosov
author
1982
text
Avtomaticheskaya Svarka [Automatic Welding]
1982
6
351
62
63
Estimating chemical composition dependence of weld joint metal radiation embrittlement.
NP
Anosov
author
1985
text
Avtomaticheskaya Svarka [Automatic Welding]
1985
10
391
66
68
Radiation resistance evaluation for 15H2NMFAA weld joint fusion zone metal using seams of variable chemical composition.
NP
Anosov
author
1990
text
Avtomaticheskaya Svarka [Automatic Welding]
1990
11
452
7
10
2010
2010
DIN EN ISO 148-1-2011 (2010) Metallic materials – Charpy pendulum impact test – Part 1: Test method (ISO 148-1:2009); German version EN ISO 148-1:2010, 35 pp.
AV
Dub
author
2016
2016
10.1007/s10512-011-9403-7
The error in determining the critical brittleness temperature dose-time relationships for RPV steel welded joints in VVER design. VANT. Ser.
AV
Dub
author
2011a
text
Safety of nuclear power plants,
2011a
30
126
141
Critical brittleness temperature dose-time relationships for predicting VVER-1000 RPV steel lifetime.
AV
Dub
author
2011b
text
Atomnaya Energiya [Atomic Energy]
2011b
110
3
123
130
1982
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GOST 5639-82 (1982) Steel and alloys. Methods for detection and determination of grain size: Russian standard. Moscow: IPC Publishing House of Standards, 21 pp. [in Russian]
1994
Metals.
1994
26 pp
GOST 9454-78 (1994) Metals.Method for testing the impact strength at low, room and high temperature: Russian standard. Moscow: IPC Publishing House of Standards, 26 pp. [in Russian]
2014
Metallic materials. Charpy pendulum impact test: Russian standard.
2014
46 pp
GOST R ISO 148-1-2013 (2014) Metallic materials. Charpy pendulum impact test: Russian standard.Moscow: Russian scientific and technical centre for information on standardization, metrology and conformity assessment, 46 pp.
AG
Kazantsev
author
2015
2015
Reference heat treatment methods for VVER type reactor vessel shells.
SI
Markov
author
2011
text
Tyazhyoloe mashinostroenie [Heavy engineering]
2011
8
2
16
1987
Rules of strength calculation for equipment and pipelines of nuclear power plants: Russian Rules and Standards in Nuclear Power Engineering. Gosatomenergonadzor of the USSR.
1987
525 pp
PNAE G-7-008-89 (1987) Rules of strength calculation for equipment and pipelines of nuclear power plants: Russian Rules and Standards in Nuclear Power Engineering. Gosatomenergonadzor of the USSR.Energoatomizdat Publ., Moscow, 525 pp.
2012
Procedure for VVER-1000 reactor vessel strength and lifetime calculation based on fracture toughness values determined by the surveillance specimens testing: Russian Operating Company Guidance Document.
2012
56 pp
RD EO 1.1.2.09.0789-2012 (2012) Procedure for VVER-1000 reactor vessel strength and lifetime calculation based on fracture toughness values determined by the surveillance specimens testing: Russian Operating Company Guidance Document.JSC “Rosenergoatom”, Moscow, 56 pp.
2013
Calculations of brittle fracture resistance for VVER vessels at the design stage: Russian Operating Company Guidance Document.
2013
63 pp
RD EO 1.1.2.99.0920-2013 (2013) Calculations of brittle fracture resistance for VVER vessels at the design stage: Russian Operating Company Guidance Document.JSC “Rosenergoatom”, Moscow, 63 pp.
AV
Surkov
author
1979
1979
2003
2003
TU 0893-013-00212179-2003 (2003) Billets of 15H2NMFA, 15H2NMFA-A or 15H2NMFA grade 1 steels for reactor plant vessels, covers, and other components: Russian Technical Standard. Information and Reference system “Teksekspert”, 26 pp.
Effect of impurities on radiation resistance of 15H2NMFA perlitic steel.
OM
Vishkarev
author
1980
text
Proceedings of TsNIITMASh
1980
157
19
24
Radiation resistance of 15H2NMFA steel.
OM
Vishkarev
author
1980a
text
Proceedings of TsNIITMASh
1980a
157
4
6
10.3897/nucet.4.30779
https://nucet.pensoft.net/article/30779/
https://nucet.pensoft.net/article/30779/download/pdf/
https://nucet.pensoft.net/article/30779/download/xml/
The authors investigate the influence of chemical and structural inhomogeneity on the brittle fracture resistance (BFR) of VVER vessel materials in the initial state (without irradiation). It is proposed to replace the brittle fracture resistance assessment using the critical brittleness temperature TC for the BFR assessment using the brittle-viscous transition temperature TT. Consideration was given to calibration charts used for studying the TT dependence on the grain size and heat treatment.
A comparison of the TC and TT values in the experimental industrial 15H2NMFA-A steel billets shows that the TC values are significantly lower than the TT values:
– at the lower level of conservatism, the difference between TC and TT is 22 °C;
– at the upper level of conservatism, this difference is 24 °C.
The array data on the critical brittleness temperature and the ductile-to-brittle transition temperature of impact test samples of 15H2NMFAA (for VVER-1000) and 15H2NMFA grade 1 (for VVER-1200) steels were statistically processed. The industrial shell samples were manufactured at the “Energomashspetsstal” plant (Kramatorsk, Ukraine).
It was found that, in the metal of VVER-1000 vessel surveillance specimens with the copper content
– less than 0.06%, heat treatment has a significant effect on the TT value, which changes from –99 to –28°C;
– from 0.07 to 0.12%, heat treatment has a significant effect on the TT value, which changes from –60 to –40°C.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Brittle fracture resistance assessment
critical brittleness temperature
ductile-to-brittle transition temperature
standard deviation
conservative estimates of the VVER-1000 vessel life.
Brittle fracture resistance of reactor pressure vessel steels in the initial state
Research Article
10.3897/nucet.4.30780
2018-12-07
nucet
LLC “Laboratory of materials of Obninsk Institute for Nuclear Power Engineering”, Obninsk, Russia
Obninsk Institute for Nuclear Power Engineering (INPE NRNU MEPhI), Obninsk, Russia
author
Stepanov, Vladimir
JSC "SSC RF – IPPE", Obninsk, Russia
author
Chernov, Vladimir
FGBUN Interdepartmental Center of Analytical Researches in the Field of Physics, Moscow, Russia
author
Parshikov, Yury
JSC “ELEKOND”, Sarapul, Russia
author
Lebedev, Viktor
JSC “ELEKOND”, Sarapul, Russia
Udmurt State University, Izhevsk, Russia
author
Kharanzhevsky, Yevgeny
2018-12-07
2018-12-07
2018
Nuclear Energy and Technology
2452-3038
4
163-166
2018
VB
Anufrienko
author
2006
2006
VB
Anufrienko
author
2008
2008
10.1016/j.cnsns.2009.05.066
10.1109/TIA.2002.804762
Production and research of properties of nanostructures for direct transformation of nuclear energy in electric with use of issue of secondary electrons.
VA
Chernov
author
2010
text
Nano- and microsystem technology
2010
2010
11
2
9
VA
Chernov
author
2011
2011
Development of nanostructured secondary electron energy converters for the creation of miniature current sources of constant readiness.
VA
Chernov
author
2015
text
Nano- and microsystem technology
2015
2015
2
57
64
Miniature nanostructured current sources based on direct conversion of nuclear energy.
VA
Chernov
author
2016
text
Russian Chemical Journal
2016
60
3
20
25
PP
Maltsev
author
2005
2005
10.1016/S0378-7753(00)00387-6
10.1109/PES.2008.4596576
10.1016/j.jpowsour.2015.03.122
10.1134/S1063784216020249
VD
Verner
author
2008
2008
10.1201/b14671
10.3897/nucet.4.30780
https://nucet.pensoft.net/article/30780/
https://nucet.pensoft.net/article/30780/download/pdf/
https://nucet.pensoft.net/article/30780/download/xml/
In current sources with a radioactive isotope (CSRI), nuclear energy is directly converted into electricity due to the separation of electric charges during the decay of radioactive isotopes. It was previously shown that asymmetric supercapacitors can be used as CSRI prototypes if, after being exposed to pulsed reactor irradiation, the electric charge on their plates increases to several coulombs as a result of internal induced activity. In this paper, the electric charge separation and accumulation in supercapacitors were studied directly in the process of neutron irradiation.
The study was focused on the electrophysical characteristics of cylindrical supercapacitors with an organic electrolyte produced by JSC “ELEKOND”. A comparison of symmetric and asymmetric supercapacitors showed that an effective charge accumulation occurs in the asymmetric capacitors: it is independent of the neutron flux density and determined by the absorbed radiation dose. The electrical voltage between the plates of a symmetrical supercapacitor with a capacity of 100 F during irradiation up to an absorbed dose of 50 Gy reaches 1.24 mV. When asymmetric supercapacitors are irradiated with the same dose, a significant increase in the potential difference up to 1.15 V is observed during irradiation and for a long time afterwards (1.5·105 s) due to the electric charge redistribution (~ 5·10–3 C) in the electrolyte and carbon particles with the formation of a double electrical layer. The post-radiation increase in the capacity of asymmetric supercapacitors is ~ 5 mF.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Supercapacitor
neutron irradiation
radiation-induced electric charge
Radiation-induced separation and accumulation of electric charge in supercapacitors
Research Article
10.3897/nucet.4.27346
2018-12-07
nucet
Missouri University of Science and Technology, Rolla, United States of America
Alexandria University, Alexandria, Egypt
author
Said, Ibrahim
Missouri University of Science and Technology, Rolla, United States of America
Alexandria University, Alexandria, Egypt
author
Moharam, Mahmoud
Missouri University of Science and Technology, Rolla, United States of America
author
Alexander, Vineet
Missouri University of Science and Technology, Rolla, United States of America
author
Usman, Shaoib
Missouri University of Science and Technology, Rolla, United States of America
author
Al-Dahhan, Muthanna
2018-12-07
2018-12-07
2018
Nuclear Energy and Technology
2452-3038
4
167-178
2018
funder
U.S. Department of Energy
10.13039/100000015
IAS
Abdallah
author
2017
2017
10.1016/j.anucene.2015.10.038
RJ
Aldridge
author
2013
2013
10.1016/j.tsep.2018.06.011
R
Aris
author
1956
1956
JA
Castañeda
author
2014
2014
10.1016/0009-2509(53)80001-1
HS
Fogler
author
1999
1999
L
Han
author
2007
2007
L
Han
author
2005
2005
10.1515/ijcre-2018-0039
10.1016/j.nucengdes.2005.10.027
10.1016/0923-0467(94)02922-9
10.1002/aic.690010306
O
Levenspiel
author
1972
1972
10.1016/0098-1354(89)85062-8
MMT
Moharam
author
2017
2017
10.1016/j.ijthermalsci.2017.10.004
10.1016/j.expthermflusci.2017.12.027
10.1002/aic.15583
10.1002/aic.690380705
10.1016/j.nucengdes.2007.03.010
MM
Taha
author
2017
2017
10.1016/j.expthermflusci.2018.02.003
10.1016/j.expthermflusci.2018.02.003
G
Taylor
author
1954
1954
10.1115/IMECE2011-64259
10.1016/j.nucengdes.2013.04.009
10.1016/j.nucengdes.2016.03.002
10.1016/j.nucengdes.2014.04.012
10.1016/0009-2509(90)85012-3
C-Y
Wen
author
1975
1975
10.3897/nucet.4.27346
https://nucet.pensoft.net/article/27346/
https://nucet.pensoft.net/article/27346/download/pdf/
https://nucet.pensoft.net/article/27346/download/xml/
Multiphase Reactors Engineering and Applications Laboratory performed gas phase dispersion experiments in a separate-effect cold-flow experimental setup for coolant flow within heated channels of the prismatic modular reactor under accident scenario using gaseous tracer technique. The separate-effect experimental setup was designed on light of local velocity measurements obtained by using hot wire anemometry. The measurements consist of pulse-response of gas tracer that is flowing through the mimicked riser channel using air as a carrier. The dispersion of the gas phase within the separate-effect riser channel was described using one-dimensional axial dispersion model. The axial dispersion coefficient and Peclet number of the coolant gas phase and their residence time distribution within were measured. Effect of heating intensities in terms of heat fluxes on the coolant gas dispersion along riser channels were mimicked in the current study by a certain range of volumetric air flow rate ranging from 0.0015 to 0.0034 m3/s which corresponding to heating intensity range from 200 to 1400 W/m2. Results confirm a reduction in the response curve spreads is achieved by increasing the volumetric air velocity (representing heating intensity). Also, the results reveal a reduction in values of axial dispersion coefficient with increasing the air volumetric flow rate.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Prismatic modular reactor
gaseous tracer technique
axial dispersion coefficient
Peclet number
Axial dispersion and mixing of coolant gas within a separate-effect prismatic modular reactor
Research Article
10.3897/nucet.4.31867
2018-12-07
nucet
JSC SSC RF-IPPE n.a. A.I. Leypunsky, Obninsk, Russia
author
Kirillov, Andrey
JSC SSC RF-IPPE n.a. A.I. Leypunsky, Obninsk, Russia
author
Yarygin, Valeriy
2018-12-07
2018-12-07
2018
Nuclear Energy and Technology
2452-3038
4
179-183
2018
2017
2017
Agilent Technologies website (2017) http://www.agitech.ru/ [Accessed Sep. 1, 2017] [In Russian]
Basic Package of Hardware, Procedural and Software Tools for Experimental Research of Laboratory TICs.
AV
Andriashin
author
1996
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Selected works of IPPE. Obninsk. IPPE Publ.
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76
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MN
Arnoldov
author
2012
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VV
Denisenko
author
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AV
Grafkin
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2017
ICP DAS websit (2017) Available at http://www.icpdas.com/[accessed Sep. 1, 2017]. [In Russian]
Intelligent thermocouple signal input modules for a thermionic experimental facility.
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Kirillov
author
2016
text
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2016
3
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Resistive current shunts for high-power applications.
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Kolpakov
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Kuznetsov
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Software for the three-dimensional numerical calculation of the thermal and electrical characteristics of the multi-cell thermionic fuel element for a thermionic NPP. Izvestiya vysshykh uchebnykh zavedeniy.
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Polous
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Pyatnitskiy
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Rannev
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Shiryaev
author
2009
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Sinyavskiy
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10.1007/BF02673518
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Nuclear power of direct energy conversion in space missions of the 21st century. Izvestiya vysshykh uchebnykh zavedeniy.
VI
Yarygin
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Yadernaya energetika,
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Experimental Studies and Testing of TICs (Instrument 0100) to Support the Characteristics of a Standardized TFE.
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Yarygin
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Yarygin
author
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10.3897/nucet.4.31867
https://nucet.pensoft.net/article/31867/
https://nucet.pensoft.net/article/31867/download/pdf/
https://nucet.pensoft.net/article/31867/download/xml/
Studies and tests are conducted to determine the performance of thermionic nuclear power plants (TNPP) a stage in which is pre-irradiation testing of laboratory thermionic converters (TIC) with flat and cylindrically shaped electrodes using test facilities fitted with automated data measurement systems (DMS). The TIC volt-ampere characteristics (VAC) are measured in the DMS jointly with the measured test section and experimental test facility temperature fields. The structure and the characteristics of a DMS based on products from ICP DAS Co., Ltd are presented. A developed VAC measurement program providing the operator with a convenient graphic interface and enabling adjustment of the measurement parameters has been considered. The VAC recording errors in the process of measurements have been determined using TIC simulators. The error in the VAC diffusion portion on a simulator (with a current of less than 3 A) is not more than 1%. Thanks to the use of modern components, the developed DMS offers extended functional capabilities for measuring the thermocouple signals in an experimental electrophysical test facility. The DMS structure provides for the convenience of scaling (through a larger number of measuring channels) and makes it possible to add modules from other manufacturers. The experience of operating this DMS will be used to develop the DMS for an in-pile test system designed for similar functions.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Thermionic converter
data measurement system
volt-ampere characteristic
thermocouples
pulse mode
A modern data measurement system to study and test thermionic heat to electricity converters
Research Article
10.3897/nucet.4.31857
2018-12-07
nucet
Nizhny Novgorod State Technical University n.a. R.E. Alekseev, Nizhny Novgorod, Russia
author
Melnikov, Vladimir
Nizhny Novgorod State Technical University n.a. R.E. Alekseev, Nizhny Novgorod, Russia
author
Ivanov, Vadim
Nizhny Novgorod State Technical University n.a. R.E. Alekseev, Nizhny Novgorod, Russia
author
Teplyashin, Ivan
Nizhny Novgorod State Technical University n.a. R.E. Alekseev, Nizhny Novgorod, Russia
author
Timonin, Mikhail
2018-12-07
2018-12-07
2018
Nuclear Energy and Technology
2452-3038
4
185-190
2018
10.1007/978-3-642-20233-9
10.1109/TIM.2007.903596
BE
Dozer
author
1968
1968
GV
Glebovich
author
1984
1984
JL
Gravel
author
2005
2005
WR
Hook
author
1998
1998
2018
2018
Instructions for installing and operating the BM100 level transmitter company KROHNE (2018) . www.ste.ru/krohne/pdf/rus/russ_op_manualBM100.pdf [accessed Mar 10 2018] [In Russian]
10.1002/hyp.513
Design of guided wave radar level meter based on equivalent time sampling. Proc. of the IEEE Int. Conf. “Communications, Circuit and Systems (ICCCAS)”. Chengdu, China, 15–17 Nov.
G
Jun
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2
139
142
The study of ultrasonic waveguide level gage of nuclear reactor coolant on the basic of reflex-radar principle. Izvestia visshikh uchebnikh zavedeniy.
VI
Melnikov
author
2015
text
Yadernaya energetika,
2015
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35
10.1109/5289.975464
R
Reimelt
author
2002
2002
NA
Tarasov
author
2018
Using the method of pulse reflectometry for determination of damage of cable lines
2018
1999
1999
Teploenergetika i teplotechnika (1999) Vol. 1. General questions. Section 7. Physico-chemical properties and technology solutions. Moscow. MEI Publ., 528 pp.
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Trenkal
author
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Vorontsov
author
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10.1109/ICMTCE.2013.6812478
10.3897/nucet.4.31857
https://nucet.pensoft.net/article/31857/
https://nucet.pensoft.net/article/31857/download/pdf/
https://nucet.pensoft.net/article/31857/download/xml/
The article considers the design of a microwave reflex-radar level gauge of the nuclear reactor coolant. The main advantage of the reflex-radar measurement principle is that it does not affect the accuracy of measuring the level of bubbles present, coolant condensation and boiling, changes in its pressure as well as temperature and density. In addition, the measuring transmitter design is quite simple.
In this level gauge, a microwave waveguide made as a coaxial line is used as a transducer (measuring probe). The probe consists of a steel pipe with an external diameter of 20 mm and a central electrode: it is located vertically and immersed in a controlled coolant. The probe wave resistance is 50 ohms. The device electrical diagram is presented. The oscillograms of the received signals and the basic relationships explaining the level gauge operation are given. The signals of the coaxial measuring probe are studied in a fluid with a variable dielectric constant. The results of an experimental study of the level gauge operation in a water coolant at high parameters are given: at pressures up to 10 MPa and temperatures up to 310 °C. It is shown that the device maintains its functional stability under these conditions. The level gauge’s readings practically need not be corrected when the coolant’s thermophysical properties change. The device is intended for use in the control and management systems of nuclear power plants as well as in fuel reprocessing plants.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Microwave reflex-radar level gauge
high pressure water coolant
measuring probe
nuclear reactor
power-generating equipment
Development and study of a microwave reflex-radar level gauge of the nuclear reactor coolant
Research Article
10.3897/nucet.4.31859
2018-12-07
nucet
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow, Russia
author
Kulikov, Yevgeny
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow, Russia
author
Kulikov, Gennady
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow, Russia
author
Apse, Vladimir
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow, Russia
author
Shmelev, Anatoly
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow, Russia
author
Geraskin, Nikolay
2018-12-07
2018-12-07
2018
Nuclear Energy and Technology
2452-3038
4
191-195
2018
2016
Available at: https://en
2016
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10.14429/dsj.63.5764
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2001
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10.13182/NT80-A32527
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10.1155/2010/409310
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G
Kessler
author
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2011
G
Kessler
author
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2008
A computational model and physical and technical factors that define the proliferation resistance of plutonium. Izvestiya vuzov.
EG
Kulikov
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Yadernaya energetika,
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Kulikov
author
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Computational models for the quantitative evaluation of proliferation resistance for fissionable materials. Izvestiya vysshykh uchebnykh zavedeny.
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Kulikov
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Massey
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10.3897/nucet.4.31859
https://nucet.pensoft.net/article/31859/
https://nucet.pensoft.net/article/31859/download/pdf/
https://nucet.pensoft.net/article/31859/download/xml/
The mathematical model presented in (Kulikov et al. 2018) can be used for the quantitative evaluation of the plutonium proliferation resistance. This requires the warm-up process of an implosion nuclear explosive device (NED) with a different structure to be analyzed with respect to various heat removal conditions and the option to be identified in which the NED remains operational for the longest time possible. The fraction of the 238Pu isotope with which, even in this case, the NED will prove to be operational only for quite a short time can be regarded as sufficient for the plutonium with such composition to be considered a proliferation resistant material.
The purpose of the paper is to evaluate in quantitative terms the content of 238Pu in plutonium for ensuring its proliferation resistance and to identify the factors which influence significantly this evaluation.
The data, procedures and findings from earlier works on the topic, as well as the authors’ own estimates and calculations were used for the study.
It has been shown that the important factors involved in the plutonium proliferation resistance evaluation are the NED technology level and the required NED lifetime.
Depending on the required lifetime, tougher requirements can be introduced with respect to the 238Pu content both from the standpoint of low-technology and high-technology NEDs.
With a lifetime of five hours taken as the guide-mark (a NED is unlikely to be finally assembled, transported and used for such a short time), it is only plutonium containing 55% of 238Pu that can be considered a proliferation resistant fissile material.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Plutonium
plutonium-238
proliferation resistance
nuclear explosive device
explosive
cryogenic temperatures
Quantitative evaluation of the plutonium proliferation resistance
Research Article
10.3897/nucet.4.31862
2018-12-07
nucet
Obninsk Institute for Nuclear Power Engineering National Research Nuclear University “MEPhI”, Obninsk, Russia
author
Belyavtsev, Ivan
Obninsk Institute for Nuclear Power Engineering National Research Nuclear University “MEPhI”, Obninsk, Russia
author
Starkov, Sergey
2018-12-07
2018-12-07
2018
Nuclear Energy and Technology
2452-3038
4
197-201
2018
M
Abadi
author
2015
TensorFlow: Large-scale machine learning on heterogeneous systems
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Filatova
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10.3897/nucet.4.31862
https://nucet.pensoft.net/article/31862/
https://nucet.pensoft.net/article/31862/download/pdf/
https://nucet.pensoft.net/article/31862/download/xml/
The WWR-c reactor reactivity margin can be calculated using a precision reactor model. The precision model based on the Monte Carlo method (Kolesov et al. 2011) is not well suited for operational calculations. The article describes the work on creating a software package for preliminary evaluations of the WWR-c reactor reactivity margin.
The research has confirmed the possibility of using an artificial neural network to approximate the reactivity margin based on the reactor core condition. Computational experiments were conducted on training the artificial neural network using the precision model data and real reactor measured data. According to the results of the computational experiments, the maximum relative approximation error ∆k/k for fuel burnup was 3.13 and 3.56%, respectively. The mean computation time was 100 ms.
The computational experiments showed it possible to construct the artificial neural network architecture. This architecture became the basis for building a software package for evaluating the WWR-c reactor reactivity margin – REST API based web-application – which has a convenient user interface for entering the core configuration. It is also possible to replenish the training sample with new measurements and train the artificial neuron network once again.
The reactivity margin evaluation software is ready to be tested by the WWR-c reactor personnel and to be used as a component of the automated reactor refueling system. With minor modifications, the software package can be used for reactors of other types.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
WWR-c reactor
reactivity margin
artificial neural networks
software package
generalized approximation theorem
Reactivity margin evaluation software for WWR-c reactor
Research Article
10.3897/nucet.4.31863
2018-12-07
nucet
Nuclear Safety Institute of the Russian Academy of Science, Moscow, Russia
author
Seleznev, Evgeny
Nuclear Safety Institute of the Russian Academy of Science, Moscow, Russia
author
Bereznev, Valery
2018-12-07
2018-12-07
2018
Nuclear Energy and Technology
2452-3038
4
203-209
2018
Testing of the MCU-FR code as applied to calculation of criticality of fast reactors. VANT. Ser.
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Alekseev
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10.3897/nucet.4.31863
https://nucet.pensoft.net/article/31863/
https://nucet.pensoft.net/article/31863/download/pdf/
https://nucet.pensoft.net/article/31863/download/xml/
The importance of calculation of radiation fields inside in-reactor cavities is associated with the necessity to simulate the emergency modes in fast breeder reactors (FBR), as well as reactor states with different coolant levels in special dedicated channels of passive feedback devices in lead-cooled fast reactors (LFR) of BREST type or in sodium cavities in sodium-cooled fast reactors (SFR).
The Last Flight (LF) method (Bell and Glesston 1974, Davison 1960, DOORS3.2 1988, Mynatt et. al. 1969, Rhoades and Childs 1988, Rhoades and Sipmson 1997, SCALE 2009, Voloschenko et. al. 2012), or the method of the unscattered component is widely known and is commonly used in computer codes based on the method of spherical harmonics for obtaining solution in a gas medium at a certain distance from the calculated volume domain (DORT (Rhoades and Childs 1988), TORT (Rhoades and Sipmson 1997) and others (SCALE 2009)). The practice of its application (DOORS3.2 1988) demonstrated that acceptable results are achievable at considerable distances from the surface separating dense and gas media (more than two meters). Obtaining high-quality solution is not guaranteed for cavities within the calculation area.
In addition, it is desirable to implement the cavities calculation methodology within the framework of the approximations used in reactor calculations introducing certain specific features. In particular isotropy of the neutron flux density and the necessity of forced introduction of a “conditional” calculation cell on the boundary surface of the void cavity are assumed in the diffusion approximation. If the LF method is oriented on the connection of the source point with the detection point, then it is necessary to determine in the calculation of neutron field in the cavities the neutrons escaping the surface area of the source and neutrons reaching a certain surface area of the cavity. In order to solve the problem, the authors suggested using the approximate solution presented in the paper.
Thus, an algorithm for calculation of in-reactor cavities using the diffusion approximation was developed and implemented by the authors.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Fast breeder reactor
diffusion approximation
cavities calculations.
Application of diffusion approximation in the calculations of reactor with cavities
Research Article
10.3897/nucet.4.31865
2018-12-07
nucet
“Sosny” R&D Company, Dimitrovgrad, Russia
author
Kanashov, Boris
“Sosny” R&D Company, Dimitrovgrad, Russia
author
Smirnov, Valery
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow, Russia
author
Kadilin, Vladimir
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow, Russia
author
Ibragimov, Renat
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow, Russia
author
Dedenko, Grigory
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow, Russia
author
Vlasik, Konstantin
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow, Russia
author
Rudenko, Vladimir
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow, Russia
author
Glagovsky, Eduard
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow, Russia
author
Lupar, Yevgeny
JSC NIITFA, Moscow, Russia
author
Poletov, Grigory
JSC NIITFA, Moscow, Russia
author
Lomtev, Yevgeny
JSC “The Institute in Physical - Technical Problems”, Dubna, Russia
author
Smirnov, Aleksandr
JSC “The Institute in Physical - Technical Problems”, Dubna, Russia
author
Khrunov, Vladimir
2018-12-07
2018-12-07
2018
Nuclear Energy and Technology
2452-3038
4
211-215
2018
YuK
Akimov
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2014
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Akimov
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Bovin
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Bushuev
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Dmitrenko
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10.1016/S0969-8043(99)00238-9
VV
Levenets
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2018
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Levenets
author
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Gamma and alpha spectrometry by semiconductor detectors based on CdTe (CdZnTe) produced by NSC KIPT
2018a
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Medvedev
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Reilly
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Van Loef
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10.3897/nucet.4.31865
https://nucet.pensoft.net/article/31865/
https://nucet.pensoft.net/article/31865/download/pdf/
https://nucet.pensoft.net/article/31865/download/xml/
The paper reports the first test results for detectors of various types and equipment of gamma-spectrometry channels under external radiation originating from pyrochemical reprocessing of spent mixed nitride uranium-plutonium (MNUP) fuel. Testing was carried out on a solid-state detector with a CdZnTe crystal, a scintillation detector with a LaBr3crystal, and an ionization chamber based on compressed xenon. Simulated external gamma-radiation was created by means of a Co-based scattered gamma-radiation source. The paper also describes an experimental facility and a measurement technique, and presents the facility testing results for the above three detectors. The solid-state detector was proved to have the best performance. However, achieving the design characteristics of the gamma-spectrometry channel requires new solutions for protection and collimation of gamma-radiation produced by a real MNUP SNF reprocessing facility. What is meant here is the influence of the detectors’ geometry on the configuration of the protective collimator which is proposed to be used in real conditions. Thus, if a Xe-based detector is used, the calculated mass of the protective collimator is 900 kg, while it is possible to use less massive protection for the other detectors. In addition, when manufacturing neutron shielding for detectors based on CdZnTe and LaBr3, it is necessary to consider the neutron radiation factor in MNUP SNF processing. It is possible to surround the collimator with a moderating layer (for example, polyethylene) and create inside it a skin from a thermal neutron absorber (for example, based on cadmium).
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Pyrochemical process
mixed nitride uranium-plutonium spent nuclear fuel (MUPN SNF)
pyroelectrochemical refining
fission products
on-line monitoring of nuclear materials
gamma-spectrometry
experimental facility
Capabilities of gamma-spectrometry methods for on-line monitoring of nitride SNF pyrochemical reprocessing
Research Article
10.3897/nucet.4.31861
2018-12-07
nucet
JSC "SSC RF – IPPE named after A.I. Leypunsky", Obninsk, Russia
author
Zherdev, Gennady
JSC SSC RF – IPPE named after A.I. Leypunsky, Obninsk, Russia
author
Kislitsyna, Tamara
JSC SSC RF – IPPE named after A.I. Leypunsky, Obninsk, Russia
author
Nikolayev, Mark
2018-12-07
2018-12-07
2018
Nuclear Energy and Technology
2452-3038
4
217-222
2018
SM
Bednyakov
author
2018
2018
AA
Blyskavka
author
2014a
2014a
AA
Blyskavka
author
2014b
2014b
DE
Cullen
author
1974
1974
TS
Kislitsina
author
2013
2013
TS
Kislitsina
author
2016
2016
GN
Manturov
author
1999
1999
1987
1987
MCNP (1987) A General Monte Carlo N-Particle Transport Code. Version 5 / Volume I. Overview and Theory. LANL LA-UR-03-1987.
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LM
Petrie
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Zhemchugov
author
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2018
10.3897/nucet.4.30662
GM
Zherdev
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Zherdev
author
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2015
GM
Zherdev
author
2003
2003
10.3897/nucet.4.31861
https://nucet.pensoft.net/article/31861/
https://nucet.pensoft.net/article/31861/download/pdf/
https://nucet.pensoft.net/article/31861/download/xml/
Results of studies aimed at the further refinement of the ROCOCO system (routine for calculation and organization of combined constants including cross-sections in group and subgroup representation with detailed description of energy dependence of neutron cross-sections) (Zherdev et al. 2018, Kislitsina and Nikolaev 2016) are presented in the paper. Inclusion of this system as a physical module into a set of Monte Carlo calculation codes with OOBG geometric module from the MMK code (Zherdev et al. 2003) is discussed. OOBG module is designed for calculation of neutron multiplication systems with heterogenous cores arranged as hexagonal grids with different degrees of complexity. The name ROCOCO-MMK was assigned to the complex. Results of testing the complex in the calculations of multi-zone neutron multiplication systems (including those with zones containing neutron moderator, zones with close composition but with different temperature, etc.) are described. Accounting for the dependence of constants for one and the same nuclide in the zones with different compositions and temperatures required substantial modernization of routines for preparation of constants for calculation described in (Zherdev et al. 2018). Algorithm for preparation of subgroup constants was modified, methodology for taking into account resonance self-screening of cross-sections within the range of unresolved resonances was improved, and other changes were introduced in the process of this modernization.
Results of calculations are compared with data obtained using the MCNP-5 precision program (MCNP 1987), which is linked to the same library of evaluated neutron data ROSFOND as that used in ROCOCO. The ROCOCO-MMK includes procedures for registering different neutron flux functionals (also based on ROCOCO data), which allowed including it in the SCALA computation complex (Zherdev et al. 2003, Zherdev 2005), and performing step-by-step calculation of evolution of fuel nuclide composition during the fuel residence campaign. Directions for further development of the system are outlined in conclusion and, in particular, some possibilities of using the created software for further improvement of methods for preparation of few-group constants for calculations in diffusion approximation are examined.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
New ROCOCO combined system of neutron constant preparation
further refinement
introduction in the practice of calculations
Monte-Carlo method
results of comparative calculations
potential of development
ROCOCO system of combined neutron constants – current status and results of testing using geometrical module of the MMK code
Research Article
10.3897/nucet.4.31871
2018-12-13
nucet
Obninsk Institute for Nuclear Power Engineering (INPE NRNU MEPhI), Obninsk, Russia
author
Sobolev, Artem
Obninsk Institute for Nuclear Power Engineering (INPE NRNU MEPhI), Obninsk, Russia
author
Danilov, Pavel
Obninsk Institute for Nuclear Power Engineering (INPE NRNU MEPhI), Obninsk, Russia
author
Zevyakin, Aleksandr
OON NTP “DIP”, St. Petersburg, Russia
author
Kurkov, Sergej
2018-12-13
2018-12-13
2018
Nuclear Energy and Technology
2452-3038
4
243-249
2018
2010
U.S.
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10.3897/nucet.4.31871
https://nucet.pensoft.net/article/31871/
https://nucet.pensoft.net/article/31871/download/pdf/
https://nucet.pensoft.net/article/31871/download/xml/
Results of calculation seismic resistance analysis of light equipment of nuclear power plants performed on the example of a ventilation unit using two most common analytical techniques - linear spectral analysis and direct dynamic methods - are discussed.
The basic concepts, assumptions and limitations of the linear spectral method are described. Examples are given of specific calculation cases when the method in question is not applicable in the generally accepted formulation. In particular, the phase difference and, possibly, accelerations (displacements) must be taken into consideration in the calculations of extended spatial structures for mutually remote boundary conditions. Another example are the reservoirs not completely filled with liquids. In such case waves may be formed in the liquid and taking them into account is not possible in the linear spectral method.
Specific features are examined of application of the dynamic analysis method including the input data, approaches and methodologies required for synthesizing the calculated accelerograms. A sequence of operations performed during synthesizing calculated accelerograms is provided, materials are provided containing the description of the mathematical apparatus applied for deriving the final mathematical relations for calculating response spectra and the calculation relations as such are given. The concept of the damping coefficient is explained, its influence on the calculated results and the approaches to its determination are demonstrated. Options with complete absence of damping and with absolute damping are discussed.
A real ventilation set applied in active ventilation systems of nuclear power plants was accepted as the test model. Results calculated for the detailed finite-element model of the ventilation unit using the Zenith-95 software package are presented. These results include the distribution of the calculated reduced stresses. Analysis of the results obtained using the two methods demonstrated overestimation of calculated results by the linear spectral method as compared to those obtained by the dynamic analysis method, which means that the former method underestimates the equipment’s resistance to seismic effects. In addition, the dynamic method shows additional areas in the ventilation unit where significant reduced stresses are found while the linear spectral method ignores these areas.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Seismic stability
linear spectral method
dynamic analysis method
reduced stresses
accelerogram
finite element model
Comparison of two key analysis methods for the seismic stability of equipment on the example of a ventilation unit
Research Article
10.3897/nucet.4.31873
2018-12-13
nucet
Ural Federal University, Yekaterinburg, Russia
author
Hossain, Ismail
Bangladesh Atomic Energy Commission, Dhaka, Bangladesh
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Akbar, Mohammad
Ural Federal University, Yekaterinburg, Russia
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Velkin, Vladimir
Ural Federal University, Yekaterinburg, Russia
author
Shcheklein, Sergey
2018-12-13
2018-12-13
2018
Nuclear Energy and Technology
2452-3038
4
251-256
2018
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Alam
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10.3897/nucet.4.31873
https://nucet.pensoft.net/article/31873/
https://nucet.pensoft.net/article/31873/download/pdf/
https://nucet.pensoft.net/article/31873/download/xml/
Bangladesh lies in a tectonically active zone. Earlier geological studies show that Bangladesh and its adjoining areas are exposed to a threat of severe earthquakes. Earthquakes may have disastrous consequences for a densely populated country. This dictates the need for a detailed analysis of the situation prior to the construction of nuclear power plant as required by the IAEA standards. This study reveals the correlation between seismic acceleration and potential damage. Procedures are presented for investigating the seismic hazard within the future NPP construction area. It has been shown that the obtained values of the earthquake’s peak ground acceleration are at the level below the design basis earthquake (DBE) level and will not lead to nuclear power plant malfunctions. For the most severe among the recorded and closely located earthquake centers (Madhupur) the intensity of seismic impacts on the nuclear power plant site does not exceed eight points on the MSK-64 scale. The existing predictions as to the possibility of a super-earthquake with magnitude in excess of nine points on the Richter scale to take place on the territory of the country indicate the necessity to develop an additional efficient seismic diagnostics system and to switch nuclear power plants in good time to passive heat removal mode as stipulated by the WWER 3+ design. A conclusion is made that accounting for the predicted seismic impacts in excess of the historically recorded levels should be achieved by the establishment of an additional efficient seismic diagnostics system and by timely switching the nuclear power plants to passive heat removal mode with reliable isolation of the reactor core and spent nuclear fuel pools.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Seismic hazard parameters
IAEA
seismic acceleration and damage
peak ground acceleration
Seismic safety evaluation during site selection for the nuclear power plants in Bangladesh
Research Article
10.3897/nucet.4.31875
2018-12-13
nucet
National Research Center “Kurchatov Institute", Moscow, Russia
author
Bylkin, Boris
National Research Moscow State University of Civil Engineering, Moscow, Russia
author
Engovatov, Igor
National Research Center “Kurchatov Institute, Moscow, Russia
author
Kozhevnikov, Alexey
Joint Stock Company “State Specialized Design Institute”, Moscow, Russia
author
Sinyushin, Dmitry
2018-12-13
2018-12-13
2018
Nuclear Energy and Technology
2452-3038
4
257-262
2018
10.1007/978-94-009-5628-5_32
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Decree of the RF Government No. 1069 of October 19 (2012) On Criteria of Referring Solid, Liquid and Gaseous Wastes to “Radioactive Wastes Category; Criteria of Referring Radioactive Wastes to “Special Radioactive Wastes” Category and to “Disposed Radioactive Wastes” Category; and Criteria for Classification of “Disposed Radioactive Wastes”. [“Collected Legislation of the Russian Federation”, no. 44, Art. 6017; 09.02.2015, No. 6, Art. 974].
Volume of radioactive waste and activation radiation shielding of nuclear installations.
IA
Engovatov
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2011
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Vestnik MSGU,
2011
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325
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Optimization of the Compositions of Concrete for Radiation protection of NPP.
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Engovatov
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Engovatov
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Engovatov
author
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2005
208 pp
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Engovatov
author
1999
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Minimization of radioactive wastes in decommissioning of new generation nuclear power plants.
IA
Engovatov
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51
Long-lived activation products in Light-Water Reactor Construction Materials: Implication for Decommissioning.
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John Evans
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Radioactive Waste Management and the Nuclear Fuel Cycle,
1988
11
1
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39
Radionuclide vector.
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Ivanov
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Rosenergoatom,
2015
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42
45
Activation studies of concrete binding agent ingredients used for nuclear radiation shielding.
VM
Nazarov
author
1991
text
Kernenergie,
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34
7
8
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NUREG/CR-200 (1995) Rev. 5. SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation.
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1999
ONL-ESICC DLC-185 (1999) Bugle-96: Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications.
1998
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ORNL-RSICC C-650 (1998) DOORS 3.2: One-, Two- and Three Dimensional Discrete Ordinates Neutron/Photon Transport Code System.
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1999
Ser. No. WS-G-2.1 (1999) IAEA Safety Standards. Decommissioning of Nuclear Power Plants and Research Reactors. Vienna, 41 pp.
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Voytkevich
author
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1977
10.3897/nucet.4.31875
https://nucet.pensoft.net/article/31875/
https://nucet.pensoft.net/article/31875/download/pdf/
https://nucet.pensoft.net/article/31875/download/xml/
Existing situation in nuclear industry is characterized with simultaneous development of the following two processes: design and construction of new generation of nuclear installations and decommissioning of installations of older generations.
Significant amounts of radioactive wastes generated during the decommissioning phase are determined both for the first and the second types of installations by the induced activity of neutron irradiated structural and shielding materials. Concentration of the so-called radioactivity-hazardous nuclides in primary building and construction materials is the most important characteristics determining the resulting levels of induced activity. Values of these concentrations for the same type of material extracted from different geological deposits may differ by one or two orders of magnitude. Information about concentrations of radiation-hazardous elements in radiation shielding materials is fragmented and, as a rule, unsuitable for practical application.
The purpose of the present study was to substantiate the necessity of compiling and recording the data on the concentrations of radioactivity-hazardous nuclides for building and structural materials for nuclear installations during the phases of design, operation and decommissioning.
Three types of shielding concrete compositions were selected for the investigation. Concentrations of radioactivity-hazardous nuclides were mainly obtained by neutron activation technique. Neutron transport calculations were performed in one-dimensional cylindrical geometry at the core mid-plane according to usual core-vessel-shielding model of modern VVER reactor unit including 2-m thick concrete shield. Both transport and activation calculations were undertaken using modules of SCALE system.
The obtained results allow estimating neutron-induced activation levels in the material as the function of irradiation time, amounts and categories of radioactive waste and their evolution during the decay time from 1 to 100 years. It was established that neutron-induced activity of shielding concrete strongly depends on the actual concentrations of radioactivity-hazardous nuclides in the concrete including ‘trace’ concentrations (other factors being the same). It was also shown that failure to take such concentrations into account may lead to the underestimation of neutron-induced activation levels and amounts of radioactive wastes and their category.
The obtained results confirmed the necessity of compiling and maintaining data records on the concentrations of radioactivity-hazardous nuclides for materials used in structural and shielding materials of nuclear installations. Proposals were formulated on the potential accumulation of information, composition and formatting of descriptors of chemical composition of shielding and structural materials of nuclear installations.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Nuclear facilities
NPP unit
decommissioning
neutron induced activity
radioactive waste
radiation shielding concrete
descriptor
On the necessity and the role of descriptors of neutron activated structural and shielding materials of nuclear installations for future decommissioning
Research Article
10.3897/nucet.4.31889
2018-12-13
nucet
State Research Center of the Russian Federation “Troitsk Institute for Innovation & Fusion Research”, Moscow, Russia
author
Kalinichev, Peter
State Research Center of the Russian Federation “Troitsk Institute for Innovation & Fusion Research”, Moscow, Russia
author
Evdokimov, Igor
State Research Center of the Russian Federation “Troitsk Institute for Innovation & Fusion Research”, Moscow, Russia
author
Likhanskii, Vladimir
2018-12-13
2018-12-13
2018
Nuclear Energy and Technology
2452-3038
4
263-270
2018
CE
Beyer
author
1991
1991
DL
Burman
author
1991
1991
10.1016/j.jnucmat.2010.01.006
V
Likhanskiy
author
2004
2004
10.2172/7343826
Lena
Oliver
author
2017
2017
Failed rod diagnosis and primary circuit contamination level determination, thanks to the DIADEME code.
D
Parrat
author
2003
text
IAEA-TECDOC-
2003
1345
265
276
Fuel assemblies of WWER-1000 nuclear reactors.
2016
text
Standard procedure for the fuel cladding leak detection with Amend. 2. Moscow. JSC Concern Rosenergoatom Publ.
2016
28
34
G
Rossiter
author
2002
2002
Ukrainian WWER-type NPP units. Results of cladding tightness inspection.
NYu
Shumkova
author
2003
text
IAEA-TECDOC-
2003
1345
77
86
Fuel failure diagnostics in normal operation of nuclear power plants with WWER-type reactors.
P
Slavyagin
author
2003
text
IAEA-TECDOC-
2003
1345
303
315
Regulation of the fission product activity in the primary coolant and assessment of defective fuel rod characteristics in steady state WWER-type reactor operation.
P
Slavyagin
author
2003a
text
IAEA-TECDOC-
2003a
1345
326
337
JNDC nuclear data library of fission products, version 2.
Kanj
Tasaka
author
1990
text
JAERI,
1990
1320
92
96
2018
2018
The IAEA nuclear energy series (2018) Review of fuel failures in water cooled reactors in 2006-2015: 43-45.
10.1016/0022-3115(82)90419-6
10.1016/S0022-3115(01)00571-2
10.1016/0022-3115(85)90028-5
10.1016/0022-3115(88)90316-9
10.3897/nucet.4.31889
https://nucet.pensoft.net/article/31889/
https://nucet.pensoft.net/article/31889/download/pdf/
https://nucet.pensoft.net/article/31889/download/xml/
Fuel failures during operation of Nuclear Power Plants (NPPs) may lead to substantial economic losses. Negative effects of reactor operation with leaking fuel in the core may be reduced if fuel failures are detected in due time of the cycle.
At present time, the ratio of the normalized release rates of 131I and 134I is used to detect fuel failures in WWERs during steady state operation. However, based on the activity of iodine radionuclides, it is not always possible to detect the fuel failure. This situation may occur in case of a small defect in cladding of a leaking fuel rod or for high burnup fuel if the defect is overlapped by the surface of the fuel pellet. If it is so, fuel deposits may be the dominant contributor to iodine activity, and the fuel failure may not be noticeable.
In PWRs, fuel failures are detected by activity of radioactive noble gases. Noble gases are not adsorbed on cladding inner surface, as distinct from iodine radionuclides. Release of noble gases from the leaking fuel rod may be considerable even though defect in cladding is small.
In this paper, a technique is proposed for detection of fuel failures at WWER reactors by activity of radioactive noble gases in the primary coolant. It is shown that radioactive noble gases may be a more sensitive indicator of fuel failures than iodine radionuclides. Detection of fuel failures is based on monitoring of the ratio between 133Xe and 135Xe activity. Some examples of practical applications are given.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
WWER
fuel rod
fuel failure
fission products
technique
coolant activity
iodine radionuclides
radioactive noble gases
A technique for detection of WWER fuel failures by activity of Xe radionuclides during reactor operation
Research Article
10.3897/nucet.4.31891
2018-12-13
nucet
JSC “SSC RF-IPPE n.a. A.I. Leypunsky”, Obninsk, Russia
author
Usanov, Vladimir
2018-12-13
2018-12-13
2018
Nuclear Energy and Technology
2452-3038
4
271-277
2018
E Waltar
Alan
author
2012
Fast Spectrum Reactors.
2012
720 pp
LS
Belyaev
author
2009
2009
IA
Blank
author
2010
2010
2007
2007
GIЕ/EMWG /2007/ 004. Revision 4.2 (2007) Cost estimating guidelines for generation IV nuclear energy systems, 181 pp.
2013
2013
IAEA Nuclear Energy Series NP-T-1.14 (2013) Framework for Assessing Dynamic Nuclear Energy Systems for Sustainability, Final Report of the INPRO Collaborative Project GAINS: 252.
2014
INPRO Methodology for sustainability assessment of nuclear energy systems: Economics. IAEA Nuclear Energy Series No. NG-T-4.4.
2014
92 pp
INPRO Manual (2014) INPRO Methodology for sustainability assessment of nuclear energy systems: Economics. IAEA Nuclear Energy Series No. NG-T-4.4.IAEA, Vienna, 92 pp.
AN
Karkhov
author
1998
1998
AN
Karkhov
author
1999
1999
Development prospects of nuclear power in market conditions.
AN
Karkhov
author
2014
text
Problemy prognozirovaniya,
2014
4
26
37
OD
Kazachkovskiy
author
2000
2000
Criteria for the return on investments in nuclear power.
VV
Kharitonov
author
2017
text
Izvestiya vuzov, Yadernaya energetika,
2017
2
157
168
VV
Kovalev
author
1995
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2015
2015
Projected costs of generating electricity (2015) International Energy Agency (NEA), N.7054, OECD, 212 pр.
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Rachkov
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2008
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YaV
Shevelev
author
1996
1996
Methodological problems of application of the weighted average cost of capital (WACC) in financial calculations.
AK
Solodov
author
2013
text
Finansovy menedzher,
2013
3
35
41
2000
Есonomiс Еvaluation of Bids for Nuсlear Power Plants. 1999 Edition.
2000
224 pp
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2004
2004
The Economic Future of Nuclear Power (2004) University of Chicago, 368 pp.
1994
1994
The Economics of Nuclear Fuel Cycle (1994) Nuclear Energy Agency. Organization for Economic Development and Cooperation. Paris, 164 pp.
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The Future of Nuclear Power (2003) An Interdisciplinary MIT Study, 170 pp.
10.3897/nucet.4.31891
https://nucet.pensoft.net/article/31891/
https://nucet.pensoft.net/article/31891/download/pdf/
https://nucet.pensoft.net/article/31891/download/xml/
Possibilities are analyzed for improving the commercial attractiveness of nuclear electricity generation in market conditions. A model is presented in which a financially integrated electricity generating system comprising several units of one technological type, rather than a single unit, is subject to an economic analysis. Issues have been considered involved in the calculation of the electricity cost in such systems and their construction. It has been shown that the calculated unit cost of the electricity generated in a financially integrated nuclear energy system with the number of units being more than one, provided it is financed by shareholders and creditors, can be lower as compared with the cost of the electricity generated by power units, not integrated economically, of the same capacity under the same investment conditions.
The effect is achieved thanks to the short-term crediting component in the electricity cost the funds on which can be returned, at a time, for a smaller number of units (even for only one), as electricity is produced by all units in the system. The results of the calculations for nuclear energy sources and combined-cycle plants using the developed model make it possible to conclude that the switch from economic models of individual nuclear units to models of integrated energy systems can bring the calculated economic performance of nuclear power closer to (or better) the performance of fossil-fuel energy sources. If achieved, this may increase the commercial attractiveness of nuclear power and contribute to a growth in the public and private investments in nuclear power business.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Nuclear power business
competitiveness
electricity cost
capital market
funding mechanisms
computational model
financially integrated system
Systemic competitiveness of nuclear energy sources
Research Article
10.3897/nucet.4.31892
2018-12-13
nucet
Experimental Scientific Research and Methodology Center “Simulation Systems” (SSL), Obninsk, Russia
author
Dorovskikh, Vyacheslav
Experimental Scientific Research and Methodology Center “Simulation Systems” (SSL), Obninsk, Russia
author
Dorokhovich, Sergey
Experimental Scientific Research and Methodology Center “Simulation Systems” (SSL), Obninsk, Russia
author
Zajtsev, Aleksey
Experimental Scientific Research and Methodology Center “Simulation Systems” (SSL), Obninsk, Russia
author
Levchenko, Valery
Experimental Scientific Research and Methodology Center “Simulation Systems” (SSL), Obninsk, Russia
author
Leonov, Igor
2018-12-13
2018-12-13
2018
Nuclear Energy and Technology
2452-3038
4
279-285
2018
AM
Bubenchikov
author
2001
2001
RO
Gauntt
author
2001
MELCOR Computer Code Manuals, NUREG/CR-6119, SAND2001-0929P.
2001
231 pp
10.1016/j.nucengdes.2015.09.013
2001
2001
IAEA-TECDOC-1196 (2001) International Atomic Energy Agency (IAEA), Mitigation of hydrogen hazards in water cooled power reactors. Vienna, 42 pp.
IA
Kirillov
author
2017
2017
2002
2002
KUPOL-M code. Ver. 1.1 (2002) Program description. Report No. 82022/4. Obninsk. FEI Publ., 42 pp. [in Russian]
LD
Landau
author
1988
1988
RE
Launder
author
1976
1976
VA
Levchenko
author
2007
2007
KK
Murata
author
1997
Code Manual for CONTAIN 2.0: A Computer Code for Nuclear Reactor Containment Analysis, Rep. NUREG/CR-6533, Rep.
1997
388 pp
S
Patankar
author
1984
1984
10.1016/j.fusengdes.2015.03.021
AA
Samarsky
author
1978
1978
ZM
Shapiro
author
1957
Hydrogen Flammability Data and Application to PWR LOCA, WAPD-SC-545, Westinghouse Electric Corp.
1957
23 pp
On ways to reduce the risk of fires in turbine halls of nuclear power plants.
GE
Soldatov
author
2009
text
Atomcon,
2009
3
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1994
1994
Test M-7-1 (1994) NUPEC Hydrogen Mixing and Distribution Test, Final Comparison Report on ISP-35, NEA/CSNI/R(94)29, 109 pp.
L
Valencia
author
1992
Design Report, Hydrogen Distribution Experiments El1.1-E11.
1992
95 pp
AA
Zaitsev
author
2005
2005
10.3897/nucet.4.31892
https://nucet.pensoft.net/article/31892/
https://nucet.pensoft.net/article/31892/download/pdf/
https://nucet.pensoft.net/article/31892/download/xml/
The article gives a general description of the SIMCO calculation code designed to simulate thermohydraulic and physico-chemical processes in containments of nuclear power facilities. The authors present a calculation technique based on a physico-mathematical model in lumped parameters. As a numerical solution method, the modified semi-implicit SIMPLER procedure is used. The code was examined using analytical and qualitative tests. A comparison of the numerical and analytical solutions showed good agreement. The code was verified using the experimental data obtained at the NUPEC installation (Japan). Based on the results of testing and verification, it was concluded that, in general, physico-mathematical code models adequately describe the processes of heat/mass transfer in the containment. Therefore, this SIMCO code version can be used to analyze the totality of thermophysical and physico-chemical processes in nuclear power facilities with containments, including the transfer of hydrogen/steam/air mixtures.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
SIMCO code
physico-mathematical model
lumped parameters
analytical test
verification
hydrogen distribution
The SIMCO containment code applied to modeling hydrogen distribution in containments of nuclear power facilities
Research Article
10.3897/nucet.4.30379
2018-12-13
nucet
Higher Technological Institute, Ramadan city, Egypt
author
Galahom, Ahmed
https://orcid.org/0000-0002-1280-9363
2018-12-13
2018-12-13
2018
Nuclear Energy and Technology
2452-3038
4
287-293
2018
10.1016/j.anucene.2008.11.006
10.1016/j.nucengdes.2010.10.036
10.1016/j.anucene.2014.10.025
10.1016/j.anucene.2016.02.025
JS
Hendricks
author
2008
2008
Design studies of a typical PWR core using advanced computational tools and techniques.
MQ
Huda
author
2011
text
Annals of Nuclear Energy,
2011
38
1939
1949
10.1016/j.nucengdes.2010.06.026
AL
Nichols
author
2008
2008
PM
O’Leary
author
2000
2000
CE
Sanders
author
2002
2002
MS
Yahya
author
2014
2014
A
Yamamoto
author
2002
2002
10.3897/nucet.4.30379
https://nucet.pensoft.net/article/30379/
https://nucet.pensoft.net/article/30379/download/pdf/
https://nucet.pensoft.net/article/30379/download/xml/
This article examines the effect of an integral burnable absorber (IBA) on the neutronic characteristics of Pressurized Water Reactor (PWR) to provide possible improvements for the fuel management. MCNPX code was used to design a three dimensional model for PWR assembly. The designed model has been validated by comparing the output data with a previously published data. MCNPX code was used to analyze the radial thermal neutron flux and the radial power distribution through PWR assembly with and without IBA. Gadolinium is burnable absorber material that was used in the IBA rods. The gadolinium element suppressed the power in the regions where they were distributed. The existence of IBA rods has a large effect on the Kinf. This effect decreases gradually with burnup due to the degradation of gadolinium. The gadolinium isotopes degradation was analyzed with burnup. Different numbers of IBA rods were investigated to optimize the suitable number that can be used in the PWR assembly. The gadolinium effect on the concentration of 135Xe and 149Sm resulting from the fission process was analyzed.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
PWR
IBAs
MCNPX code
Gd2O3
Power
Thermal neutron flux
Simulate the effect of integral burnable absorber on the neutronic characteristics of a PWR assembly
Research Article
10.3897/nucet.4.31868
2018-12-13
nucet
FSUE VNIIA, Moscow, Russia
author
Belonosov, Mikhail
FSUE VNIIA, Moscow, Russia
NRNU MEPhI, Moscow, Russia
author
Kishkin, Vladimir
NRNU MEPhI, Moscow, Russia
author
Korolev, Sergey A.
2018-12-13
2018-12-13
2018
Nuclear Energy and Technology
2452-3038
4
223-228
2018
DE
Baburin
author
2018
2018
The end-to-end engineering tools for instrumentation and control systems for nuclear power plants.
MA
Belonosov
author
2015
text
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2015
2
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51
System engineering of the automated process control system of NPPs.
OL
Bozhenkov
author
2009
text
Yadernye izmeritelno-informacionnye tehnologii [Nuclear Engineering and Information Technology],
2009
2
27
30
MA
Yastrebenetsky
author
2011
2011
PCS of power units of nuclear power plants with VVER.
VG
Dunaev
author
2011
text
In: Nuclear Power. Problems. Solutions. Part 1. (Ed.) Strikhanov MN. Moscow: TsSPiM Publ.
2011
356
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Structural synthesis of automation schemes in conditions of incomplete requirements for technical implementation.
NN
Filatova
author
2012
text
Izvestiya VolGTU [Bulletin of VolSTU],
2012
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22
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IAEA NS-G-1.1 (2000) The software of control systems, important for safety, executed on the basis of computer equipment. Safety Guide. Vienna. IAEA.
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2012
IAEA SSR-2/1 (2012) Safety of nuclear power plants: design. Specific safety requirements. Vienna. IAEA.
2002
2002
IEC 61513-2002 (2002) Nuclear power plants. Monitoring and control systems important for safety. General requirements.
10.1088/1742-6596/781/1/012048
Modern methods of verification of software and hardware complexes of automated process control systems of nuclear power plants based on TPTS.
SA
Korolev
author
2016
text
Elektricheskie stantsii [Electric Power Plants],
2016
8
9
15
H
Miedl
author
1996
1996
The complex of automation system TPTS-SB.
AD
Naritz
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text
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author
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10.1109/SWAT.1971.10
The structure of the automated process control system of the Belarusian NPP in terms of safety.
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Timohin
author
2015
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2
88
28
32
VV
Zverkov
author
2014
2014
Analysis of approaches to the construction of automated process control systems of NPPs.
VV
Zverkov
author
2015
text
Elektricheskie stantsii [Electric Power Plants],
2015
8
2
6
Program-Technical Complexes of Control Systems for Safety of Nuclear Power Plants.
VV
Zverkov
author
2017
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Elektricheskie stantsii,
2017
1
2
10
VE
Zyubin
author
2005
2005
10.3897/nucet.4.31868
https://nucet.pensoft.net/article/31868/
https://nucet.pensoft.net/article/31868/download/pdf/
https://nucet.pensoft.net/article/31868/download/xml/
The article describes an automated verification method used for application software of control safety systems based on the TPTS-SB equipment. Verification is performed by comparing two mathematical models (oriented graphs): one obtained by processing the original design data, i.e., graphical functional diagrams, and the other formed by reversing the program code loaded from the controller. The vertices in both graphs are functional blocks of mathematical and logical operations; the edges are connections between them. The constructed mathematical models undergo a comparison, covering the vertices and edges of the graphs as well as the memory cells and values of constants. The equivalence of mathematical models proves the correspondence between the program code and the initial set of design functional diagrams.
The proposed automated verification method makes it possible to prove that no distortion is introduced into the program during the process of converting graphical functional diagrams into the program code with its subsequent translation and loading into the controller. It is postulated that any distortions will be detected during the verification procedure, which is performed every time after loading the code into the controller.
The solution provides an acceptable speed when large volumes of vector graphics stored in a relational database are processed, and makes it possible to visualize the verification results. The proposed method is implemented in the GET-R1 instrumentation tools for TPTS-SB and is used in designing and verifying the application software of the safety systems at the Belarusian NPP.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Verification
reverse engineering
code generation
safety systems
controller
mathematical model
instrumentation tools
Verification on application program generation and loading for safety systems of nuclear power plants based on the reverse engineering method
Research Article
10.3897/nucet.4.31869
2018-12-13
nucet
NRNU MEPhI, Moscow, Russia
author
Andrianov, Andrey
NRNU MEPhI, Moscow, Russia
author
Korovin, Yury
NRNU MEPhI, Moscow, Russia
author
Kuptsov, Ilya
Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen, Germany
author
Konobeyev, Aleksandr
JSC “SSC RF-IPPE n.a. A.I. Leypunsky", Obninsk, Russia
author
Andrianova, Olga
2018-12-13
2018-12-13
2018
Nuclear Energy and Technology
2452-3038
4
229-234
2018
10.1016/S0168-9002(03)01368-8
A
Andrianov
author
2016
2016
10.1051/epjconf/201714612007
Multi-criteria comparative evaluation of spallation reaction models.
AA
Andrianov
author
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Vestnik natsionalnogo issledovatelskogo yadernogo universiteta MIFI,
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Andrianov
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Andrianov
author
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Konobeyev
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10.1007/s10669-016-9598-1
10.3897/nucet.4.31869
https://nucet.pensoft.net/article/31869/
https://nucet.pensoft.net/article/31869/download/pdf/
https://nucet.pensoft.net/article/31869/download/xml/
The paper presents the results of a comparative evaluation of the predictive ability of seventeen spallation reaction models (CEM02, CEM03, Phits/jam, Cascade/ASF, Phits/Bertini, Bertini/Dresner, Cascade-4, INCL4/Abla, INCL4/smm, geant4/binary, Isabela/smm, geant4/Bertini, Isabela/Abla, INCL4/Gemini, CASCADeX-1.2, Isabel/Gemini, Phits/jqmd) for the interaction reactions of high-energy protons with natPb nuclei using the most popular methods of multiple-criteria decision analysis (MAVT/MAUT, AHP, TOPSIS, PROMETHEE). Multiple-criteria decision analysis methods are used extensively to support decision-making in various fields of knowledge, including nuclear physics and engineering, when aggregating conflicting criteria with due account for the expert and decision-maker opinions. Four factors of computational and experimental agreement (R, D, F, H), most commonly used in this field of knowledge, have been employed as the criteria, which, having been aggregated as part of applying respective multiple-criteria decision analysis methods, make it possible to estimate the integral measure of the computational model effectiveness and to rank the models, using this as the basis, depending on the degree of their predictive ability. It has been demonstrated that the ranking results obtained using different multiple-criteria decision analysis methods show a good agreement. Using a stochastic approach to the generation of weights, the models were ranked in conditions with the absence of data on the significance of individual agreement factors. Recommendations are presented for using the multiple-criteria decision analysis methods to address tasks involved in the preparation of nuclear data in conditions of a multiple-factor evaluation of discrepancies between calculations and experiment.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Нigh-energy nuclear reactions
nuclear data
multiple-criteria decision analysis methods
uncertainty
Comparison of spallation reaction models based on multiple-criteria decision analysis
Research Article
10.3897/nucet.4.31870
2018-12-13
nucet
Nuclear Safety Institute of the Russian Academy of Sciences, Moscow, Russia
author
Sarkisov, Ashot
Nuclear Safety Institute of the Russian Academy of Sciences, Moscow, Russia
author
Antipov, Sergey
Nuclear Safety Institute of the Russian Academy of Sciences, Moscow, Russia
author
Smolentsev, Dmitry
Nuclear Safety Institute of the Russian Academy of Sciences, Moscow, Russia
author
Bilashenko, Vyacheslav P.
Nuclear Safety Institute of the Russian Academy of Sciences, Moscow, Russia
author
Kobrinsky, Mikhail N.
Nuclear Safety Institute of the Russian Academy of Sciences, Moscow, Russia
author
Sotnikov, Vladimir A.
Nuclear Safety Institute of the Russian Academy of Sciences, Moscow, Russia
author
Shvedov, Pavel A.
2018-12-13
2018-12-13
2018
Nuclear Energy and Technology
2452-3038
4
235-241
2018
2013
2013
Approaches for assessing the economic competitiveness of small and medium sized reactors (2013) Vienna: Intern. Atomic Energy Agency, 271 pp.
2018
Available at: http://tass
2018
Compact Nuclear Batteries for Arctic are to be Constructed for Ministry of Defense by 2023 (2018) Available at: http://tass.ru/armiya-i-opk/4508435 [Accessed Feb. 7, 2018]. [In Russian]
Eddy-resolving 1/10° Model of the World Ocean. Izvestiya Rossiyskoy akademii nauk.
RA
Ibrayev
author
2012
text
Fizika atmosfery i okeana,
2012
48
1
45
55
2018
2018
JSC “Afrikantov OKBM” (2018) Reactor Facilities for Nuclear Icebreakers and Optimized Floating Blocks. Available at: http://www.okbm.nnov.ru/images/pdf/ritm-200_extended_ru_web.pdf [Accessed Feb. 7, 2018] [In Russian]
The prospects for development of nuclear-powered icebreaker fleet.
MM
Kashka
author
2016
text
Arctic: ecology and economy
2016
3
23
98
107
IV
Kudinovich
author
2016
2016
Aspects of Liability Insurance of Nuclear Risks from the Low-power Nuclear power plants. Izvestiya Rossiyskoy akademii nauk.
VP
Kuznetsov
author
2014
text
Energetika,
2014
2
88
95
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2015
Low-Power Nuclear Power Plants – A New Line in the Development of Power Systems (2015) Vol. 2. Ed. by Acad. RAN A.A. Sarkisov. Moscow. Akadem-Print Publ., 387 pp. [In Russian]
Nuclear Energy in the Arctic Region.
VS
Nikitin
author
2015
text
Arctic: ecology and economy,
2015
4
20
86
95
Results of the Scientific and Technical Board of the Unified Energy System and the Section on the Problems of Reliability and Safety of Large Energy Systems of the Scientific Council of the Russian Academy of Sciences on Electric Power Systems Research.
2016
text
Vesti v Elektroenergetike,
2016
6
86
33
44
Ole
Reistad
author
2006
2006
Rostekhnadzor Order of 09.04.2017 No. 351. Available at: htpps://rg.ru/2017/09/29/rostekhnadzor-prikaz351-site-dok.html [Accessed Feb. 7, 2018]. [In Russian]
Icebreaker support for the largest national Arctic hydrocarbon projects.
VV
Ruksha
author
2016
text
Arctic: ecology and economy,
2016
4
24
109
113
Introductory Paper of the Chairman of the Conference Scientific Committee. Low-Power Nuclear Power Plants - A New Line in the Development of Power Systems. Moscow. Nauka Publ.
AA
Sarkisov
author
2011
text
,
2011
1
7
12
10.1007/s10512-018-0385-6
Economic Efficiency and Possibilities of Using Megawatt-class Nuclear Power Sources in the Arctic.
AA
Sarkisov
author
2018a
text
Arctic: ecology and economy,
2018a
1
29
4
14
AA
Sarkisov
author
2015
2015
YuV
Sivintsev
author
2005
2005
2016
2016
Small Modular Reactors: Nuclear Energy Market Potential for Near-Term Development (2016) [S.l.]: OECD. [NEA No. 7213]
Development of the Arctic energy sector: problems and capabilities of low-power generation.
DO
Smolentsev
author
2012
text
Arctic: ecology and economy,
2012
3
7
22
29
2004
iss.
2004
168 pp
The History of Nuclear Power of the Soviet Union and Russia: A Collection of Articles (2004) iss.1–5, iss. 5: History of Low-Power Nuclear Power Plants. Ed. by V.A. Sidorenko. Ros. Nauch. Tsentr “Kurchatovsky Institut”. Moscow. IzdAT Publ., 168 pp. [In Russian]
10.3897/nucet.4.31870
https://nucet.pensoft.net/article/31870/
https://nucet.pensoft.net/article/31870/download/pdf/
https://nucet.pensoft.net/article/31870/download/xml/
The demands for nuclear power technologies in the Arctic for solving social and economic problems of the state can only be satisfied if adequate strategies of their safe handling at all stages from design to decommissioning are defined, methodological approaches and mathematical models for predicting and minimizing adverse environmental impacts of potential emergency situations at such facilities are developed, and scientifically-based results yielded within a decision-making support system for the elimination of such emergencies are applied. Special relevance of these requirements is determined by unique features of the Arctic nature and its role in the generation of climatic and hydrological processes in the World Ocean.
Main results and generalized conclusions based on the analysis of radiological consequences of the large-scale application of nuclear power industry for the benefit of economic development of the Arctic region are provided in the present paper. The analysis was performed within the framework of the complex research project “Development of the methodological approaches and mathematical models to access the environmental impact of the possible accidents at the floating nuclear power objects, model calculations of the radiation propagation in the Arctic aquatic territories in case of emergency situations”. The increasing demand for the low-power nuclear power plants for the benefit of development of remote areas, the technological and economic advantages of such power plants as well as minimal possible environmental consequences of the hypothetic accidents resulted in the qualitative changes in the attitude towards their usage. Estimation was made of the scale of application of nuclear power and results were obtained of numerical modeling of distribution of reactivity in case of accidents. The conclusion was drawn on the necessity to adhere to the low-power nuclear energy generation development strategy based on the modular design concept.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Arctic region
nuclear icebreaking fleet
Low-Power Nuclear Power Plant
development forecast
radiation safety
sea areas
mathematical modeling
Safe development of nuclear power technologies in the Arctic: prospects and approaches
Research Article
10.3897/nucet.5.33976
2019-03-20
nucet
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow, Russia
author
Hashlamoun, Taha Mohd
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow, Russia
author
Vygovsky, Sergey
Obninsk Institute for Nuclear Power Engineering, NRNU «MEPhI», Obninsk, Russia
author
Leskin, Sergey
Obninsk Institute for Nuclear Power Engineering, NRNU «MEPhI», Obninsk, Russia
author
Duman, A. Safa
2019-03-20
2019-03-20
2019
Nuclear Energy and Technology
2452-3038
5
1
9-15
2019
AYu
Anokhin
author
2001
2001
10.1023/A:1011369130707
OA
Budnikova
author
2004
2004
BA
Dementiev
author
1990
1990
TM
Hashlamoun
author
2018
2018
VV
Kharitonov
author
2007
2007
ST
Leskin
author
2011
2011
VB
Malygin
author
2001
2001
AM
Mastepanov
author
2009
2009
VL
Molchanov
author
2009
2009
1999
1999
OECD (1999) The Economics of the Nuclear Fuel Cycle. Nuclear Energy Agency. Moscow. Energoizdat Publ., 141 pp. [in Russian]
Allowance for the fluence of fast neutrons on VVER shells and test specimens for the subsequent prediction of the radiation resource of the hulls.
1999
text
Vestnik Gosatomnadzora Rossii,
1999
1
5
2
14
2001
2001
RB-018-01 (2001) Method of neutron control on the external surface of the hulls of light water power reactors of nuclear power plants. Vestnik Gosatomnadzora Rossii, 6(19): 32–47. [approved on 17.12.2001]. [in Russian]
10.13182/NSE88-A23547
Proximity to the scenario. Prospects for evolutionary development of VVER fuel.
Yu
Semchenkov
author
2011
text
Atomnaya energetika Rossii,
2011
10
25
29
S
Tomas
author
2005
2005
SB
Vygovsky
author
2011
2011
SB
Vygovsky
author
2013
2013
Development of approaches to sensitivity analysis of the neutron fluence calculation model for VVER reactors on the basis of DOORS code complex.
SE
Yanovsky
author
2011
text
Yadernaya i radiatsionnaya bezopasnost’,
2011
3
51
38
43
DG
Zhimerin
author
1978
1978
10.3897/nucet.5.33976
https://nucet.pensoft.net/article/33976/
https://nucet.pensoft.net/article/33976/download/pdf/
https://nucet.pensoft.net/article/33976/download/xml/
This article presents the results of research, that were focused on determining the optimal parameters of the extension of (reactor life-time) reactor fuel cycle in order to reduce the total operating costs of nuclear power plants during the transition from 12-month reactor fuel cycle to 18-month fuel cycle.
The relevance of the research is related to the fact that, in recent years, there is a transition at all operating nuclear power plants VVER-1000 (1200) from 12-month reactor fuel cycle to extended 18-month fuel cycle. At the same time, represent the interests to solve the problem of conservation the extension of reactor life-time while reducing the number of loaded fuel assemblies with fresh fuel assemblies, which would reduce the total operating, and fuel costs. Search for solutions of this problem is associated with mandatory implementation of all requirements for the safe operation of the reactor and the reduction of the maximum fast neutron fluence on the reactor vessel in comparison with its value at the operating nuclear power plants.
In the present work, with using the program PROSTOR software complex researched the neutron-physical characteristics of the core at the nominal parameters of the VVER-1200 reactor through the implementation of various fuel cycle strategies. The article developed various schemes of fuel-reloading for an 18-month fuel cycle with a different number of fuel assemblies. The article carries out a comparative analysis of the main parameters in the core for fuel-reloading schemes options of an 18- and 12-month fuel cycle with each other. Determine the minimum amount of fuel assemblies and provide the necessary duration of the reactor life-time for 18-month fuel cycle with using the extension of reactor life-time by reducing the power at the end of the reactor cycle to 70% of the nominal power. In the article, the arrangements of fuel assemblies were developed to provide limitations of local power by volume of the core, which reduce the fluence of fast neutrons on the reactor vessel in comparison with the projected value of the fluence. This article shows that the 18-month fuel cycle for the VVER-1200 reactor is more economical than the 12-month fuel cycle. These studies were carried out for the VVER-1200 reactor at the power of 100% of the nominal.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
PROSTOR
fluence
VVER-1000
fuel-reloading scheme
18-month fuel cycle
effective days
natural days
fuel enrichment
burn-up
operating costs
reactor vessel
Determination of 18-month fuel cycle parameters for the purpose of fuel costs minimization at the basis of use constructions of fuel assemblies in VVER-1200 reactors
Research Article
10.3897/nucet.5.33977
2019-03-20
nucet
National Research Nuclear University, Moscow, Russia
author
Naumov, Valery
2019-03-20
2019-03-20
2019
Nuclear Energy and Technology
2452-3038
5
1
17-22
2019
Тhe concept of possible involvement of thorium in the nuclear power sector. Izvestia Vysshikh Uchebnykh Zavedeniy.
PN
Alekseev
author
1999
text
Yadernaya Energetika,
1999
1
10
18
AN
Baraboshkin
author
1976
1976
The structure of uranium dioxide precipitation obtained by electrolysis of mixed molten potassium, lithium and uranium chlorides.
AN
Baraboshkin
author
1971
text
Trudy Instituta Electrokhimii UNC AN SSSR, Sverdlovsk,
1971
17
108
117
The Integral Fast Reactor. Nucl.
YI
Chang
author
1989
text
Technol,
1989
88
129
161
170
Reaction of Thorium and ThCl4 with UO2 and (Th, U)O2 in Fused Chloride Salts. J.
P
Chiotti
author
1975
text
Less-Common Metals,
1975
42
1
141
161
Investigation of molten salt systems based on thorium fluoride. Communication 1.
VS
Emelyanov
author
1956
text
Atomnaya energiya,
1956
1
4
107
112
Investigation of molten salt systems based on thorium fluoride. Communication 2.
VS
Emelyanov
author
1956a
text
Atomnaya energiya,
1956a
1
5
80
85
T
Inoue
author
1997
1997
U-232 and the Proliferation Resistance of U-233 in Spent Fuel.
J
Kang
author
2001
text
Science and Global Security,
2001
9
1
32
RB
Kotelnikov
author
1978
1978
Formation of bivalent thorium in the molten potassium chloride environment.
VYa
Kudyakov
author
1968
text
Atomnaya energiya,
1968
24
4
448
452
The equilibrium of thorium metal with melts of alkali metal chlorides containing thorium ions.
VYa
Kudyakov
author
1972
text
Trudy Instituta Electrokhimii UNC AN SSSR, Sverdlovsk,
1972
18
27
32
VM
Murogov
author
1983
1983
OV
Skiba
author
1993
1993
Electrochemical behavior of thorium in sodium chloride and equimolar mixture of sodium and potassium chlorides.
MV
Smirnov
author
1970
text
Atomnaya energiya,
1970
27
4
419
423
Study of physical, chemical and electrochemical behavior of thorium in melts of alkali metal halides.
MV
Smirnov
author
1976
text
Radiokhimiya,
1976
18
4
639
647
The Volatility of Uranium and Thorium Tetrachloride from their Molten Mixtures with Alkali Metal Chlorides.
MV
Smirnov
author
1978
text
/ Second All-Union Conference on Uranium Chemistry (16–18 October, Moscow). Abstracts of reports. Moscow. Nauka Publ.
1978
94
.
Equilibrium potentials of metals in molten electrolytes. 1. Equilibrium potentials of thorium in chloride melts.
MV
Smirnov
author
1959
text
Izvestiya AN SSSR, Otdelenie khimicheskikh nauk,
1959
2
251
258
Interaction of thorium tetrachloride with alkali metal chlorides.
AN
Vokhmyakov
author
1973
text
Atomnaya energiya,
1973
35
6
424
423
NM
Voronov
author
1971
1971
10.3897/nucet.5.33977
https://nucet.pensoft.net/article/33977/
https://nucet.pensoft.net/article/33977/download/pdf/
https://nucet.pensoft.net/article/33977/download/xml/
The use of thorium in combination with plutonium in nuclear power generation offers a solution to the problem of reducing the accumulation of long-lived transplutonium nuclides. Along with this, the existing uranium fuel cycle (UFC) has such disadvantage as the vulnerability to unauthorized use of nuclear materials. The thorium fuel cycle (TFC) is devoid of these drawbacks.
The engagement of thorium in nuclear power is possible provided the availability of an appropriate technology for reprocessing irradiated thorium. A fuel cycle based on thorium oxide may not differ in principle from the already developed pyrochemical fuel cycle involving uranium and plutonium oxides. Thorium oxide is most commonly obtained in compact state by electrolysis of molten salts from thorium-containing electrolytes. The most thorough studies of physical and chemical and electrochemical behavior of thorium in molten haloids of alkali and alkaline-earth metals were conducted in the 1960ies and the 1970ies.
Since extensive experimental material has been accumulated by now for justification of the use of pyroelectrochemical and chemical processes for regeneration of fuel in molten salts, then it has also been proposed that technologies for fuel reprocessing in molten chlorides of alkali metals should be applied resulting in a crystalline product that can be used for the fuel element fabrication.
Unlike uranium and plutonium, thorium behavior in molten salt environments is less complex. In molten salts, thorium exists predominantly in the form of Th4+, and the mixture of uranium and thorium dioxides with ThO2 content reaching up to 50 % can be obtained by electrolysis of molten salts.
Therefore, the existing amount of knowledge about the chemistry of thorium allows regarding the use of pyrochemical processes in production of thorium oxide as highly promising, and the available data on the physical and chemical properties of thorium and its compounds in high-temperature molten salts makes it possible to state that the pyroelectrochemical technology can be potentially used in production and reprocessing of thorium fuel.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Thorium
oxide fuel
thorium fuel cycle
options
pyroelectrochemical processes
electrochemical behavior of thorium
fuel reprocessing
Conceptual potential of a pyroelectrochemical technology for the thorium engagement in the fast neutron fuel cycle
Research Article
10.3897/nucet.5.33978
2019-03-20
nucet
National Research Tomsk Polytechnic University, Tomsk, Russia
author
Bedenko, Sergey
https://orcid.org/0000-0003-4318-6338
National Research Tomsk Polytechnic University, Tomsk, Russia
author
Knyshev, Vladimir
National Research Tomsk Polytechnic University, Tomsk, Russia
author
Kuznetsova, Mariya
National Research Tomsk Polytechnic University, Tomsk, Russia
author
Lutsik, Igor
National Research Tomsk Polytechnic University, Tomsk, Russia
author
Shamanin, Igor
2019-03-20
2019-03-20
2019
Nuclear Energy and Technology
2452-3038
5
1
23-29
2019
10.3897/nucet.5.33978
https://nucet.pensoft.net/article/33978/
https://nucet.pensoft.net/article/33978/download/pdf/
https://nucet.pensoft.net/article/33978/download/xml/
A computational study has been performed for various options of the thorium reactor core loading. Neutronic studies of fuel have been conducted, its isotopic composition has been calculated, and the alpha emitters and the sources of neutron and photon radiation in the microencapsulated nuclear fuel have been analyzed. The studies had the purpose of developing the methodology used to estimate the radiation characteristics of nuclear fuel with a complex inner structure. Emphasis is placed on calculating the quantitative and spectral composition of the neutrons formed as the result of (a, n) reactions on small- and average-mass nuclei. The ratio of the quantity of the neutrons resulting from the (a, n) reactions to the quantity of the neutrons formed as the result of spontaneous fission has been calculated for fuel with heterogeneous and homogeneous arrangements of fissionable and structural elements. The developed tools will make it possible to estimate the neutron radiation dose, to revise the traditional fresh and spent fuel handling procedures, and to estimate, using the Rossi alpha method, the neutron multiplication factor in deeply subcritical systems. The neutron yield and spectrum were calculated using an analytical model and verified codes such as WIMS-D5B, ORIGEN-APP, SOURCES-4C and SRIM-2013.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Thorium reactor
isotopic composition
alpha particle transport
neutron source
spectrum of radiation sources
Peculiarities of the radiation formation in dispersed microencapsulated nuclear fuel
Research Article
10.3897/nucet.5.33980
2019-03-20
nucet
Peter the Great St. Petersburg Polytechnic University, Saint Petersburg, Russia
author
Egorov, Mikhail
2019-03-20
2019-03-20
2019
Nuclear Energy and Technology
2452-3038
5
1
31-38
2019
10.3897/nucet.5.33980
https://nucet.pensoft.net/article/33980/
https://nucet.pensoft.net/article/33980/download/pdf/
https://nucet.pensoft.net/article/33980/download/xml/
Steam generators for NPPs are the important large-sized metal consuming equipment of nuclear power installations. Efficiency of steam generator operation determines the overall service life of the whole nuclear facility.
The main aim of the current study is to analyze advantages and shortcomings of horizontal and vertical types of steam generator design. This analysis is aimed at the development of recommendations for designing advanced steam generators for future Russian units of NPPs with VVER reactors of increased power.
Design solutions and fifty-year experience of operation of 400 steam generators of horizontal type accepted in Russia and of vertical type applied by Westinghouse, Combustion Engineering, Siemens, Mitsubishi, Doosan were analyzed within the framework of the present study. Advantages and drawbacks of both types of equipment determining the development of conditions of the operating processes were also identified and systematized.
Currently NPPs equipped with VVER are characterized with extended surface area of containment shells due to the application of four-loop design configuration and horizontal-type steam generators. It was established that steam generator equipment of horizontal type is characterized by such inherent disadvantages of design, technological and operational nature as the following: 1) small height and volume of the vapor space above the evaporation surface reducing separation capabilities and the capacity of the equipment as a whole; 2) impossibility of organizing separate single-phase pre-boiling section. Because of the above, horizontal steam generators with dimensions permissible for railroad transportation and, for VVER-1200 with reactor vessel diameter equal to 5 m, by water transport as well, have exhausted the possibilities for further significant increase of the per unit electric power.
The demonstrated advantages of vertical-type steam generators were as follows: 1) absence of stagnant zones within the second cooling circuit, and, consequently, of hold-ups in them; 2) uniformity of heat absorption efficiency of the heating surface ensuring, as well, improved conditions for moisture separation; 3) high degree of moisture removal from steam-water mixture due to the combination of moisture separating elements of chevron and swirl-vane types; 4) increased temperature drop with parameters of generated steam elevated by 0.3 – 0.4 MPa.
Conclusion was made on the advisability of introduction of steam generators with vertical-type layout in the Russian nuclear power generation. Practical tasks that need to be addressed in order to ensure introduction of vertical steam generators at NPPs with high-power VVER reactors were formulated.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Vertical type steam generator
horizontal type steam generator
light water nuclear reactor
natural circulation
variable cyclic thermal stresses
pre-boiling section
heat exchange
steam-water mixture
temperature difference
separators of chevron and swirl-vane type
containment
the four-loop layout.
Vertical steam generators for VVER NPPs
Research Article
10.3897/nucet.5.33981
2019-03-20
nucet
National Research Nuclear University, Obninsk, Russia
author
Andrianov, Andrey
National Research Nuclear University, Obninsk, Russia
author
Kuptsov, Ilya
National Research Nuclear University, Obninsk, Russia
author
Osipova, Tatyana
JSC “SSC RF-IPPE n.a. A.I. Leypunsky”, Obninsk, Russia
author
Andrianova, Olga
JSC “Engineering Center of Nuclear Containers”, Moscow, Russia
author
Utyanskaya, Tatyana
2019-03-20
2019-03-20
2019
Nuclear Energy and Technology
2452-3038
5
1
39-45
2019
10.3897/nucet.5.33981
https://nucet.pensoft.net/article/33981/
https://nucet.pensoft.net/article/33981/download/pdf/
https://nucet.pensoft.net/article/33981/download/xml/
The article presents a description and some illustrative results of the application of two optimization models for a two-component nuclear energy system consisting of thermal and fast reactors in a closed nuclear fuel cycle. These models correspond to two possible options of developing Russian nuclear energy system, which are discussed in the expert community: (1) thermal and fast reactors utilizing uranium and mixed oxide fuel, (2) thermal reactors utilizing uranium oxide fuel and fast reactors utilizing mixed nitride uranium-plutonium fuel. The optimization models elaborated using the IAEA MESSAGE energy planning tool make it possible not only to optimize the nuclear energy system structure according to the economic criterion, taking into account resource and infrastructural constraints, but also to be used as a basis for developing multi-objective, stochastic and robust optimization models of a two-component nuclear energy system. These models were elaborated in full compliance with the recommendations of the IAEA’s PESS and INPRO sections, regarding the specification of nuclear energy systems in MESSAGE. The study is based on publications of experts from NRC “Kurchatov Institute”, JSC “SSC RF-IPPE”, ITCP “Proryv”, JSC “NIKIET”. The presented results demonstrate the characteristic structural features of a two-component nuclear energy system for conservative assumptions in order to illustrate the capabilities of the developed optimization models. Consideration is also given to the economic feasibility of a technologically diversified nuclear energy structure providing the possibility of forming on its base a robust system in the future. It has been demonstrated that given the current uncertainties in the costs of nuclear fuel cycle services and reactor technologies, it is impossible at the moment to make a reasonable conclusion regarding the greatest attractiveness of a particular option in terms of the economic performance.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Thermal reactors
fast reactors
closed nuclear fuel cycle
UOX
MOX
MNUP
optimization
uncertainty
forecasting
MESSAGE
Optimization models of a two-component nuclear energy system with thermal and fast reactors in a closed nuclear fuel cycle
Research Article
10.3897/nucet.5.33982
2019-03-20
nucet
Obninsk Institute for Nuclear Power Engineering, Obninsk, Russia
author
Belozerov, Vladimir
Novovoronezh NPP, Branch of JSC “Concern Rosenergoatom”, Novovoronezh, Russia
author
Zhuk, Mikhail
Obninsk Institute for Nuclear Power Engineering, Obninsk, Russia
author
Terekhova, Anna
2019-03-20
2019-03-20
2019
Nuclear Energy and Technology
2452-3038
5
1
47-52
2019
10.3897/nucet.5.33982
https://nucet.pensoft.net/article/33982/
https://nucet.pensoft.net/article/33982/download/pdf/
https://nucet.pensoft.net/article/33982/download/xml/
Modes with violation of the reactor plant cooling conditions on the primary circuit side of a VVER reactor were simulated using the TRAC-PD2 and Open FOAM thermohydraulic codes (TRAC-PD2 1981, OpenFOAM User Guide Version 1.6. 2009, OpenFOAM Programmer’s Guide Version 1.6. 2009) based on energy and mass conservation equations for the three-dimensional unsteady flow of a two-phase mixture. Coupled simulation of the dynamics of neutronic and thermohydraulic processes (TRAC-PD2 1981, OpenFOAM User Guide Version 1.6. 2009, OpenFOAM Programmer’s Guide Version 1.6. 2009, Bolshagin et al. 2009, Galanin 1971, Weinberg and Wigner 1961, Ovchinnikov and Semenov 1988, Report LA-UR-03-1987) aims to improve the qualitative understanding and the quantitative presentation of their effects on safety.
Investigating these modes using the above thermohydraulic codes makes it possible to analyze the course of transients and certain emergency processes without using the industrial testing method, this providing the basis for solving the problems of ensuring the reliability, operational safety and efficiency of nuclear power plants.
A modern nuclear reactor is a complex system studying and calculating which requires more than the use of simple theoretical models. Thermohydraulic calculations are an essential part of most engineering and technological development works in nuclear power. Since, in conditions of an NPP, no technologically conventional way can be used to verify and update the results and findings of an a priori analysis on the basis of commercial tests, investigations based on codes are used in some cases as the tools to study and predict the parameters of thermohydraulic processes in the reactor’s circulation circuit.
The main purpose of the study is to calculate and investigate, based on codes, modes with violation of the reactor plant cooling conditions on the primary circuit side of a VVER reactor in order to determine if calculated parameters conform to the acceptance criteria established by regulatory documents.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Mode
pipe break
failure of heat removal
reactor plant
pressure
stop valve
departure from nucleate boiling
safety factor
high-speed dump valves with discharge to atmosphere (HSDV-A)
Investigation of the small break conditions in the primary circuit of a VVER-1000 reactor
Research Article
10.3897/nucet.5.33984
2019-03-20
nucet
Obninsk Institute for Nuclear Power Engineering, Obninsk, Russia
author
Yuferov, Anatoliy
2019-03-20
2019-03-20
2019
Nuclear Energy and Technology
2452-3038
5
1
53-59
2019
10.3897/nucet.5.33984
https://nucet.pensoft.net/article/33984/
https://nucet.pensoft.net/article/33984/download/pdf/
https://nucet.pensoft.net/article/33984/download/xml/
Issues involved in the infologic modeling of the ENDF-format nuclear data libraries for the purpose of converting ENDF files into a relational database have been considered. The transfer to a relational format will make it possible to use standard readily available tools for nuclear data processing which simplify the conversion and operation of this data array. Infological models have been described using formulas of the “Entity (List of Attributes)” type. The proposed infological formulas are based on the physical nature of data and theoretical relations. This eliminates the need for a special notation to be introduced to describe the structure and the content of data, which, in turn, facilitates the use of relational formats in codes and solution of nuclear data evaluation problems. The concept of nuclear informatics has been formulated based on relational DBMS technologies as one of the tools for solving the “big data” problem in modern science and technology. The organizational and technological grounds for the transfer of ENDF libraries to a relational format are presented. Requirements to the nuclear data presentation formats supported by relational DBMS are listed. Peculiarities of the infological model construction, conditioned by the hierarchical nature of nuclear data, are identified. The sequence for the ENDF metadata saving is presented, which can be useful for the verification and validation (testing of the structural and syntactical validity and operability) of both source data and the procedures for the conversion to a relational format. Formulas of infological models are presented for the cross sections file, the secondary neutron energy distributions file, and the nuclear reaction product energy-angle distributions file. A complete array of infological models for ENDF libraries and the generation modules of respective relational tables are available on a public website.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
ENDF libraries
infologic modeling
relational format
Infological models of the ENDF-format nuclear data
Research Article
10.3897/nucet.5.33985
2019-03-20
nucet
Sosny R&D Company, Dimitrovgrad, Russia
author
Gayazov, Artem
Sosny R&D Company, Dimitrovgrad, Russia
author
Komarov, Sergey
Sosny R&D Company, Dimitrovgrad, Russia
author
Leshchenko, Anton
Sosny R&D Company, Dimitrovgrad, Russia
author
Revenko, Ksenia
Sosny R&D Company, Dimitrovgrad, Russia
author
Smirnov, Valery
JSC «SSC RIAR», Dimitrovgrad, Russia
author
Zvir, Elena
JSC «SSC RIAR», Dimitrovgrad, Russia
author
Ilyin, Pavel
JSC «SSC RIAR», Dimitrovgrad, Russia
author
Teplov, Vadim
2019-03-20
2019-03-20
2019
Nuclear Energy and Technology
2452-3038
5
1
61-66
2019
10.3897/nucet.5.33985
https://nucet.pensoft.net/article/33985/
https://nucet.pensoft.net/article/33985/download/pdf/
https://nucet.pensoft.net/article/33985/download/xml/
The paper describes the outcomes of the experiments to study hydrogen and gaseous fission products accumulation during simulations of the wet damaged VVER-440 SNF storage in air-tight canisters with the water drained and no drying conducted. Physical and chemical processes occurring during the damaged oxide SNF storage in wet environment are discussed. The experiments were carried out in two stages: 1) preliminary soaking of fine fuel particles in water in an air-tight canister, 2) water draining and keeping the wet SNF in the air-tight canister.
The experiments were conducted one after another using the same SNF canister and differing only in the SNF soaking temperature, i.e. 25 and 80 °С.
The radionuclide release into the liquid phase during the SNF storage under water was studied. Uranium and cesium isotopic concentrations were found to reach steady values when the SNF is kept under water for more than a month. The kinetics of hydrogen and gaseous fission product accumulation in the gaseous phase during wet storage of the spent fuel in the air-tight canister with the water drained coincide for both experiments. The kinetics demonstrate an abrupt decrease of the hydrogen and gaseous fission product accumulation rate in 46 hours. The data obtained can be applied for development and verification of the damaged SNF behavior models during SNF storage in wet environment under radiolysis.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Fire and explosion safety
damaged SNF
VVER-440
radiolysis
hydrogen
gaseous fission products
UO2 dissolution
Study of hydrogen generation and radionuclide release during wet damaged oxide spent fuel storage
Research Article
10.3897/nucet.5.32239
2019-03-20
nucet
CRAMS: Center for Research in Applied Mathematics and Statistics, Beirut, Lebanon
author
Haidar, Nassar
2019-03-20
2019-03-20
2019
Nuclear Energy and Technology
2452-3038
5
1
1-7
2019
10.3897/nucet.5.32239
https://nucet.pensoft.net/article/32239/
https://nucet.pensoft.net/article/32239/download/pdf/
https://nucet.pensoft.net/article/32239/download/xml/
We demonstrate how the therapeutic utility index and the ballistic index for dynamical neutron cancer therapy (NCT) with two opposing neutron beams form a nonlinear optimization problem. In this problem, the modulation frequencies ω and ϖ of the beams and the relative time advance ε are the control variables. A Pareto optimal control vector ω* = (ω*, ϖ*, ε*) for this problem is identified and reported for the first time. The utility index is shown to be remarkably periodically discontinuous in ε, even in the neighborhood of ε*.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Accelerator Based Modulated Neutron Sources
One-Speed Neutron Diffusion
Two Opposing Neutron Beams
Dynamical NCT
Pareto Optimization.
Optimization of two opposing neutron beams parameters in dynamical (B/Gd) neutron cancer therapy
Research Article
10.3897/nucet.5.34296
2019-04-11
nucet
Obninsk Institute for Nuclear Power Engineering, National Research Nuclear University “MEPhI” 1, Obninsk, Russia
author
Pereguda, Arkady
2019-04-11
2019-04-11
2019
Nuclear Energy and Technology
2452-3038
5
1
81-87
2019
10.3897/nucet.5.34296
https://nucet.pensoft.net/article/34296/
https://nucet.pensoft.net/article/34296/download/pdf/
https://nucet.pensoft.net/article/34296/download/xml/
An analysis of statistical data of diagnostic measurements of two parameters determining the performance of the RBMK-1000 SHADR-8A flowmeters – the minimum value of the negative amplitude half-wave at the transistor flow measuring unit (TIBR) input and the mean-square deviation over the flowmeter ball rotation period – made it possible to develop a mathematical model of the flowmeter parametric reliability. This mathematical model is a random process, which is a superposition of two delayed renewal processes. Studying the flowmeter operational reliability model provides an exponential estimate of the probability that the parameters determining the flowmeter performance will not exceed the specified levels. Using the Bernoulli scheme and the probability-estimating relationship for the flowmeter performance parameters, it is possible to calculate the probability of failure-free operation of both a single reactor quadrant and the coolant flow measurement system. In addition, it becomes possible to estimate the quadrant failure rate. Important for practice is the possibility of predicting the number of failed flowmeters depending on the system operation time. An indicator of the system reliability can be the average number of failed flowmeters, the relation for which is given in the paper. All the research results were obtained without any additional assumptions about the random values distribution laws.
The obtained results can be easily generalized for the cases when the vector dimension of the determining parameters is greater than two. The use of the results of this study is illustrated by calculated quantitative values of the flowmeter parametric reliability indicators and the coolant flow measurement system.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Parametric reliability
coolant flow measurement system
random variables
time between failures
random process
mathematical time expectation
distribution function
exponential estimate
Reliability of the RBMK-1000 coolant flow measurement system
Research Article
10.3897/nucet.5.34294
2019-04-11
nucet
MPEI Branch in Volzhsky, Volzhsky, Russia
author
Kuzevanov, Vyacheslav
MPEI Branch in Volzhsky, Volzhsky, Russia
author
Podgorny, Sergey
2019-04-11
2019-04-11
2019
Nuclear Energy and Technology
2452-3038
5
1
75-80
2019
10.3897/nucet.5.34294
https://nucet.pensoft.net/article/34294/
https://nucet.pensoft.net/article/34294/download/pdf/
https://nucet.pensoft.net/article/34294/download/xml/
The need to shape reactor cores in terms of coolant flow distributions arises due to the requirements for temperature fields in the core elements (Safety guide No. NS-G-1.12. 2005, IAEA nuclear energy series No. NP-T-2.9. 2014, Specific safety requirements No. SSR-2/1 (Rev.1) 2014). However, any reactor core shaping inevitably leads to an increase in the core pressure drop and power consumption to ensure the primary coolant circulation. This naturally makes it necessary to select a shaping principle (condition) and install heat exchange intensifiers to meet the safety requirements at the lowest power consumption for the coolant pumping.
The result of shaping a nuclear reactor core with identical cooling channels can be predicted at a quality level without detailed calculations. Therefore, it is not normally difficult to select a shaping principle in this case, and detailed calculations are required only where local heat exchange intensifiers are installed.
The situation is different if a core has cooling channels of different geometries. In this case, it will be unavoidable to make a detailed calculation of the effects of shaping and heat transfer intensifiers on changes in temperature fields.
The aim of this paper is to determine changes in the maximum wall temperatures in cooling channels of high-temperature gas-cooled reactors using the combined effects of shaped coolant mass flows and heat exchange intensifiers installed into the channels. Various shaping conditions have been considered. The authors present the calculated dependences and the procedure for determining the thermal coolant parameters and maximum temperatures of heat exchange surface walls in a system of parallel cooling channels.
Variant calculations of the GT-MHR core (NRC project No. 716 2002, Vasyaev et al. 2001, Neylan et al. 1994) with cooling channels of different diameters were carried out. Distributions of coolant flows and temperatures in cooling channels under various shaping conditions were determined using local resistances and heat exchange intensifiers. Preferred options were identified that provide the lowest maximum wall temperature of the most heat-stressed channel at the lowest core pressure drop.
The calculation procedure was verified by direct comparison of the results calculated by the proposed algorithm with the CFD simulation results (ANSYS Fluent User’s Guide 2016, ANSYS Fluent. Customization Manual 2016, ANSYS Fluent. Theory Guide 2016, Shaw1992, Anderson et al. 2009, Petrila and Trif 2005, Mohammadi and Pironneau 1994).
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Core shaping
heat exchange intensification
mass flow distribution
maximum channel wall temperature
Gas-cooled nuclear reactor core shaping using heat exchange intensifiers
Research Article
10.3897/nucet.5.34293
2019-04-11
nucet
JSC «SSC RF-IPPE n.a. A.I. Leypunsky», Obninsk, Russia
author
Kirillov, Pavel
JSC «SSC RF-IPPE n.a. A.I. Leypunsky», Obninsk, Russia
author
Bogoslovskaya, Galina
2019-04-11
2019-04-11
2019
Nuclear Energy and Technology
2452-3038
5
1
67-74
2019
2018
2018
Advanced Nuclear Power Reactors (2018) Advanced Nuclear Power Reactors Generation III+ Nuclear Reactors – World Nuclear Association. http://www.world-nuclear.org [accessed Apr 10, 2018]
P
Alekseev
author
2015
2015
YD
Baranaev
author
2007
Comparative analysis of the physical characteristics of the VVER-SKD reactor with one- and two-way coolant flow patterns: IPPE Prepint-3110.
2007
21 pp
Prospects for the use of VVER-SKD in a closed fuel cycle. Izvestia Vysshikh Uchebnykh Zavedeniy.
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Baranaev
author
2015
text
Yadernaya Energetika
2015
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YD
Baranaev
author
2010
Supercritical water cooled reactor VVER-SKD – The main contender in “Super-VVER”: IPPE Preprint -3188.
2010
19 pp
RB
Duffey
author
2008
Supercritical Water-Cooled Nuclear Reactors (SCWRs): Current and Future Concepts – Steam-Cycle Options. Proc.
2008
9 pp
2014
2014
IAEA-TECDOC, Ser. No.1746 (2014) Heat Transfer Behavior and Thermo hydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRS).
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ICTP–IAEA Workshop (2018) Joint ICTP–IAEA Workshop on Physics and Technology of Innovative Nuclear Energy Systems. https://www.iaea.org/events/joint-ictp-iaea-workshop-on-physics-and-technology-of-innovative-nuclear-energy-systems [accessed Apr 10, 2018]
Prospects for the Development of an Innovative Water-Cooled Nuclear Reactor for Supercritical Parameters of Coolant.
SG
Kalyakin
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2014
text
Teploenergetika
2014
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PL
Kirillov
author
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2013
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Makhin
author
2017
2017
Conceptual proposals on VVER-SCP reactor prototype test facility. VANT. Ser.
VM
Makhin
author
2014
text
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Pioro
author
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Pioro
author
2016
2016
Nuclear power – the basis of future electricity production. Izvestia Vysshikh Uchebnykh Zavedeniy.
IL
Pioro
author
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Yadernaya Energetika
2015
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SB
Ryzhov
author
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2011
Problems in the development of supercritical water-cooled reactor core (VVER-SKP). VANT. Ser.: Obespechenie bezopasnosti AES.
SB
Ryzhov
author
2009
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Reaktornye ustanovki s VVER-SKD
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Trukhny
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2010
422 pp
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2016
Uranium (2016) Resources, Production and Demand. / A Joint Report by the Nuclear Energy Agency and the International Atomic Energy Agency. NEA No. 7301. OECD.
10.3897/nucet.5.34293
https://nucet.pensoft.net/article/34293/
https://nucet.pensoft.net/article/34293/download/pdf/
https://nucet.pensoft.net/article/34293/download/xml/
Existing conditions make possible obtaining information that being discussed openly by wide scientific community could help outlining or even establishing the expediency of a particular area of present and future research. Use link http://www.sciencedirect.com to learn about the topics or areas that most attract researchers from different countries.
The Generation IV International Forum (GIF-IV) established in January 2000 has set a goal to improve the new generation of nuclear technologies in the following areas: stability, safety and reliability, economic competitiveness, proliferation resistance and physical protection.
The purpose of the present publication is to prepare a discussion of one of the directions of development of fourth-generation NPPs, the groundwork for which has already been laid in thermal power engineering in various countries. The number of papers published annually on this topic is the largest among other similar topics dedicated to nuclear power plants of the fourth generation.
Judging from the operating experience of existing nuclear power plants using water as a coolant, it can be ascertained that the tendency of building water-cooled nuclear power plants will remain during the next 30 to 50 years. During the present stage the task in the development of alternative types of reactors will be limited to demonstration of their performance and acceptability for future power engineering and the society.
The project of supercritical water-cooled reactor is based on the operating experience of VVER, PWR, BWR reactors (more than 14,000 reactor-years); many years of experience accumulated in operating fossil thermal power plants (more than 400 power units; 20,000 years of operation of power units) using supercritical (25 MPa, 540°C) and super-supercritical (35–37 MPa, 620–700°C) water steam. In Russia more than 140 supercritical pressure units are currently in operation.
Numerical calculation and design of supercritical water-cooled reactor (similarly to BR-10 reactor) will allow not only training personnel for future development of this technology, but will also help revealing the most difficult points requiring experimental confirmation with application of independent test facilities, as well as formulating the plan of first priority experimental studies.
Knowledge accumulated over the last 10 years in the world allows the following: further specifying the already developed concept; developing a plan of specific priority studies; compiling task order for designing small-power pilot VVER SKP-30 reactor (30 MW-th).
The scope of problems that are to be solved to substantiate VVER-SCP reactor and commence designing an experimental reactor with thermal capacity of 30 MW is the same as that in developing any type of nuclear reactor: physics of the reactor core; material related matters (primarily concerned with the reactor pressure vessel, fuel, and fuel rod cladding); thermal hydraulics of rod bundles in the near- and supercritical areas; water chemistry at supercritical pressure; corrosion of materials, development of safety systems. Research must be carried out both in static conditions and under irradiation.
The absence in Russia during the extended time period of approved program with allocation of appropriate funding and preservation of the existing status during the coming two or three years will lead to the situation when Russia will be hopelessly lagging behind in the development of SCWR technology.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
GIF-IV program
supercritical water reactor (SCWR)
prospects for nuclear power development
advantages of the VVER-SKP concept
problems of developing the VVER-SKP
proposals for cooperation
number of publications.
Generation IV supercritical water-cooled nuclear reactors: Realistic prospects and research program
Research Article
10.3897/nucet.5.35373
2019-04-12
nucet
NRNU MEPhI, Moscow, Russia
author
Andrianov, Andrey
NRNU MEPhI, Moscow, Russia
author
Korovin, Yury
NRNU MEPhI, Moscow, Russia
author
Kuptsov, Ilya
Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen, Germany
author
Konobeyev, Aleksandr
JSC “SSC RF-IPPE n.a. A.I. Leypunsky”, Obninsk, Russia
author
Andrianova, Olga
2019-04-12
2019-04-12
2019
Nuclear Energy and Technology
2452-3038
5
1
89-89
2019
10.3897/nucet.5.35373
https://nucet.pensoft.net/article/35373/
https://nucet.pensoft.net/article/35373/download/pdf/
https://nucet.pensoft.net/article/35373/download/xml/
Corrigenda: Comparison of spallation reaction models based on multiple-criteria decision analysis. https://doi.org/10.3897/nucet.4.31869
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Corrigenda: Comparison of spallation reaction models based on multiple-criteria decision analysis. https://doi.org/10.3897/nucet.4.31869
Corrigendum: Comparison of spallation reaction models based on multiple-criteria decision analysis. https://doi.org/10.3897/nucet.4.31869
Corrigendum
10.3897/nucet.5.35579
2019-05-17
nucet
NRNU MEPhI, Moscow, Russia
author
Vygovsky, Sergey
NRNU MEPhI, Moscow, Russia
author
Gruzdov, Fedor
NRNU MEPhI, Moscow, Russia
author
Al Malkawi, Rashdan
2019-05-17
2019-05-17
2019
Nuclear Energy and Technology
2452-3038
5
97-102
2019
JB
Ainscough
author
1982
Gap Conduction in Zircaloy-Clad LWR Fuel Rods. Paris.
1982
52 pp
JB
Ainscough
author
1979
1979
VG
Artemov
author
2007
2007
2010
Evolutionary and innovative development of VVER facilities
2010
ATOMEXPO-2010 International Forum (2010) Evolutionary and innovative development of VVER facilities. Report by S.B. Ryzhov, Director – General Designer of OKB Gidropress. Available at: http://2010.atomexpo.ru/mediafiles/u/files/Present/7.5_ryzhov.pdf[accessed Apr 04, 2018]. [in Russian]
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Bartolomey
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1982
Fundamentals of the Theory and Calculation Methods in Nuclear Power Reactors.
1982
510 pp
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Geelhood
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FRAPCON-3,5: A Computer Code for the Calculation of Steady-State Thermal-Mechanical Behavior of Oxide Fuel Rods for High Burnup.
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152 pp
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Kudrov
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Effective Fuel Temperature of WWER-1000.
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4 pp
10.1016/S0022-3115(96)00404-7
A
Medvedev
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Fuel Rod Behaviour at High Burnup WWER Fuel Cycles.
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11 pp
10.1504/IJNEST.2007.014654
M
Rahgoshay
author
2007
2007
M
Rahgoshay
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2011
Study of the Role of Gap Conductance Coefficient of Fuel on Increasing Safety in WWER-1000 Reactors.
2011
12 pp
LS
Tong
author
1996
Thermal Analysis of Pressurized Water Reactors.
1996
748 pp
A computational study into the dependence of the VVER reactor core neutronic performance on the temperature distribution in fuel and its effects on the parameters of the xenon processes in the core.
SB
Vygovskiy
author
2016
text
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Vygovskiy
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2004
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Weinberg
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1961
Physical Theory of Nuclear Reactors.
1961
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Wiesenack
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Yousef
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10.3897/nucet.5.35579
https://nucet.pensoft.net/article/35579/
https://nucet.pensoft.net/article/35579/download/pdf/
https://nucet.pensoft.net/article/35579/download/xml/
This paper presents the results of the research to study the dependence of the VVER-1000 (1200) cores neutronic characteristics on the cladding – fuel pellet gap conductance coefficient in the process of the fuel burn-up. The purpose of the study was to determine more accurately the dependence of the cladding – fuel pellet gap conductance coefficient on the fuel burn-up as shown in the Final Safety Report for the Bushehr NPP and to determine the extent of the effects this dependence had on the spatial distribution of the neutron field, on the xenon accumulation rate, and on the kinetic and dynamic behavior of the reactor facility. The paper presents the results of calculating the parameters using which the heat engineering safety of the reactor core is monitored in the process of the fuel burn- up (for a generalized fuel load of a VVER-1000) during the transition to an 18-month nuclear fuel cycle. This paper also includes the results of a numerical research to determine the cladding – fuel gap conductance coefficient depending on the fuel burn-up. These results have shown that, in reality, the gap conductance coefficient dependence on the burn-up does not affect greatly the steady-state characteristics. At the same time, it affects to rather a great extent the xenon accumulation rate, specifically in the event of an extended fuel life. In conditions of maneuvering (load following) modes accompanied by the xenon processes in the reactor core. These facts should be into consideration in design of engineering codes, that used to support the operation of the VVER-1000 (1200) and full-scale simulators.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
VVER-1000
gap conductance coefficient
burn-up
xenon oscillations
reactivity
Doppler effect
A study into the dependence of the cladding-fuel pellet gap conductance on burn-up and the effects on the reactor core neutronic performance
Research Article
10.3897/nucet.5.35577
2019-05-17
nucet
JSC “SSC RF-IPPE n.a. A.I. Leypunsky”, Obninsk, Russia
author
Andrianova, Olga
JSC “SSC RF-IPPE n.a. A.I. Leypunsky”, Obninsk, Russia
author
Golovko, Yury
JSC “SSC RF-IPPE n.a. A.I. Leypunsky”, Obninsk, Russia
author
Manturov, Gennady
2019-05-17
2019-05-17
2019
Nuclear Energy and Technology
2452-3038
5
91-96
2019
Testing Covariance Matrices in the ABBN Data System. Izvestia vuzov.
O
Andrianova
author
2014
text
Yadernaya energetika
2014
2
109
117
Combined Use of Differential and Integral Experiments for Adjustment of Evaluated Nuclear Data. VANT. Ser.
O
Andrianova
author
2017
text
: Nuclear and Reactor Constants
2017
1
98
105
10.1016/j.nucet.2016.07.006
Verification of the ABBN-RF group constant library on model problems and specially selected benchmark experiments.
ON
Andrianova
author
2012
text
Yadernaya fizika i inzhiniring
2012
3
2
120
126
V
Doulin
author
2007
2007
V
Doulin
author
2007a
2007a
V
Doulin
author
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2007b
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Yu
Golovko
author
2012
text
Yadernaya energetika
2012
3
5
15
Application of the Undetermined Lagrangian Coefficients Method for Eliminating Contradictions in the Analysis of ICSBEP Benchmark Experiments on Criticality Safety. VANT. Ser.
Yu
Golovko
author
2017
text
: Nuclear and Reactor Constants
2017
2
52
60
Yu
Golovko
author
2008
2008
Applying the least squares method to estimate the constant error in criticality calculations of systems with plutonium.
YuYe
Golovko
author
2014
text
Yadernaya fizika i inzhiniring
2014
5
4
293
2018
2018
ICSBEP (2018) International Handbook of Evaluated Criticality Safety Benchmark Experiments. https://www.oecd-nea.org/science/wpncs/icsbep/handbook.html [Access date: July 05, 2018]
T
Ivanova
author
2012
2012
Codes and Nuclear Data for Reactor Neutronics Calculations and Uncertainty Estimation. VANT. Ser.
G
Manturov
author
2017
text
: Nuclear and Reactor Constants
2017
1
115
128
G
Manturov
author
2017
2017
2008
2008
MCNP A General Monte Carlo N-Particle Transport Code (2008) Version 5, Volume I: Overview and Theory, LA-UR-03-1987, Los Alamos, 416 pp.
2018
2018
OECD/NEA (2018) OECD/NEA web-site. https://www.oecd-nea.org/ [Access date: July 05, 2018]
2018
2018
UACSA (2018) Web-page of WPNCS Expert Group on Uncertainty Analysis for Criticality Safety Assessment (UACSA). https://www.oecd-nea.org/science/wpncs/UACSA/ [Access date: July 05, 2018]
L
Usachev
author
1972
1972
ROSFOND – Russian National Library of Evaluated Neutron Data. VANT. Ser.
SV
Zabrodskaya
author
2007
text
: Nuclear and Reactor Constants
2007
12
3
21
10.3897/nucet.5.35577
https://nucet.pensoft.net/article/35577/
https://nucet.pensoft.net/article/35577/download/pdf/
https://nucet.pensoft.net/article/35577/download/xml/
The paper presents the results of a computational analysis of the OECD/NEA benchmark conducted to estimate the accuracy of the critical safety parameters of multiplying MOX-fueled systems. The computational test is a set of 15 spherical multiplying systems that differ in their compositions and geometries. According to the test conditions, the keff values of the analyzed systems are unknown in advance. As part of the computational analysis of the test involving national codes and nuclear data libraries, along with the keff calculations, it is also necessary to estimate the a priori (due to the accuracy of the nuclear data used) and a posteriori (based on the accumulated experimental information) errors in the calculated keff values. Based on the benchmark, an updated version of the ROSFOND/ABBN-RF nuclear data was tested. The results of estimating the a priori and a posteriori errors in keff using the INDECS system for the proposed test models are presented. The analysis of the calculated data shows that (1) the observed spread in the keff values obtained from the Russian ROSFOND library and foreign evaluated nuclear data libraries (ENDF/B-VII.0, JEFF-3.2, JENDL-4.0) varies from –0.3 up to 0.8%; and (2) the deviation of the calculation results in the keff values obtained from the ROSFOND library and its group version, ABBN-RF, does not exceed 0.1%. The average a priori error in keff for all the tested options of multiplying systems is about 1% and, taking into account the selected set of experimental criticality data for MOX-fueled systems, including experiments at the BFS facilities, the average a posteriori error in keff can be reduced to 0.3%. The performed estimations confirm the high accuracy of the ROSFOND/ABBN-RF nuclear data for calculating the critical safety parameters of multiplying MOX-fueled systems.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
MOX fuel
integral experiments
BFS critical facility
accuracy estimation
effective multiplication factor
neutron data uncertainty
OECD/NEA benchmark
maximum likelihood method
Verification of the ROSFOND/ABBN nuclear data based on the OECD/NEA benchmark on criticality safety of mox-fueled systems
Research Article
10.3897/nucet.5.35797
2019-05-17
nucet
JSC “ORPE “Teсhnologiya” n.a. A.G. Romashin”, Obninsk, Russia
author
Kiryushina, Valentina
JSC “ORPE “Teсhnologiya” n.a. A.G. Romashin”, Obninsk, Russia
author
Kovaleva, Yuliya
JSC “ORPE “Teсhnologiya” n.a. A.G. Romashin”, Obninsk, Russia
author
Stepanov, Pуtr
JSC “ORPE “Teсhnologiya” n.a. A.G. Romashin”, Obninsk, Russia
author
Kovalenko, Pavel
2019-05-17
2019-05-17
2019
Nuclear Energy and Technology
2452-3038
5
103-108
2019
A
Argon
author
1978
1978
SM
Barinov
author
1996
1996
10.1115/1.2806812
VV
Bolotin
author
1965
Statistical Methods in Structural Mechanics.
1965
279 pp
VV
Bolotin
author
1980
1980
10.1177/002199837400800209
10.1016/0961-9526(94)90080-9
T
Fujii
author
1982
Fracture Mechanics of Composite Materials.
1982
232 pp
IG
Gurtovnik
author
2003
Radiotransparent Elements of Glass Reinforced Plastics.
2003
368 pp
10.1016/0010-4361(78)90590-6
GI
Ivchenko
author
1992
Mathematical Statistics: Schoolbook for Technical Universities.
1992
304 pp
10.1007/BF00540858
K
Kapur
author
1980
Reliability and Design of Systems.
1980
607 pp
Estimation of Weibull parameters in strength analysis of ceramic materials for antenna fairings.
VV
Kiryushina
author
2006
text
Mekhanika kompositnykh materialov i konstruktsiy
2006
12
1
76
82
Scale effect on the strength of glass ceramic antenna fairing.
VS
Levshanov
author
2006
text
Mekhanika kompozitsionnykh materialov i konstruktsiy
2006
12
3
312
316
Composite materials for radiotransparent fairings of aircraft.
MYu
Rusin
author
2014
text
Novye ogneupory
2014
10
8
13
Size effect in tensile test of glass reinforced plastics.
SV
Serensen
author
1962
text
Zavodskaya laboratoriya
1962
27
4
483
485
Destruction of glass reinforced plastics at short-term loading.
SV
Serensen
author
1965
text
Mekhanika polimerov
1965
2
93
103
YuM
Tarnopolsky
author
1966
Structural Strength and Deformability of Glass Reinforced Plastics.
1966
260 pp
VV
Vasilyev
author
1988
Mechanics of Structures of Composite Materials.
1988
272 pp
10.1007/BF00606005
10.3897/nucet.5.35797
https://nucet.pensoft.net/article/35797/
https://nucet.pensoft.net/article/35797/download/pdf/
https://nucet.pensoft.net/article/35797/download/xml/
Polymer composite materials (PCM) are used extensively and are viewed as candidates for application in various industries, including nuclear power. Despite a variety of methods and procedures employed to investigate the mechanical characteristics of PCMs, the use of the laboratory sample mechanical test results to design and model large-sized structures is not always fully correct and reasonable. In particular, one of the problems is concerned with taking into account the scale parameter effects on the PCM strength and elastic characteristics immediately in the product.
The purpose of the study is to investigate the scale effects on the mechanical characteristics of glass reinforced plastics using phenolformaldehyde and silicon-organic binders and a fabric quartz filler.
Samples of four different standard sizes under GOST 25604-82 and GOST 4648-2014 were tested for three-point bending using an LFM-100 test machine to estimate the scale effect. The thicknesses of the model samples were chosen with regard for the wall thicknesses of full-scale products under development or manufactured commercially and the test machine features, and varied in the limits of 1.6 to 7.5 mm.
The tests showed that strength decreased as the sample thickness was increased to 3 mm and more both at room and elevated (200 to 500 °C) temperatures, which can be described by an exponential function based on the Weibull statistical model. The values of the Weibull modulus that characterizes the extent of the scale effect on the strength of the tested materials were 4.6 to 6.7. The average bend strength in the sample thickness range of 3 mm and less does not vary notably or tends to increase slightly as the thickness is increased. This fact makes it possible to conclude that estimation of allowable stresses in a thin-wall product requires the use of test results for samples with a thickness that is equal to the product wall thickness since standard samples may yield overestimated allowable stress values and lead, accordingly, to incorrect calculations of the strength factor.
The results obtained shall be taken into account when defining the allowable levels of operation for full-scale products and structures of polymer composites based on the laboratory sample strength data as well as when estimating their robustness as a characteristic of the product’s fail-safe operation.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Polymer composite materials
glass reinforced plastics
scale effect
strength
Weibull statistical model
size effect
A study into the scale effect on the strength properties of polymer composite materials
Research Article
10.3897/nucet.5.35798
2019-05-17
nucet
SRC – Burnasyan Federal Medical Biophysical Center of Federal Medical Biological Agency, Moscow, Russia
author
Konobeev, Ivan
Obninsk Institute for Nuclear Power Engineering National Research Nuclear University MEPhI, Obninsk, Russia
author
Kurachenko, Yury
SRC – Burnasyan Federal Medical Biophysical Center of Federal Medical Biological Agency, Obninsk, Russia
author
Sheino, Igor
2019-05-17
2019-05-17
2019
Nuclear Energy and Technology
2452-3038
5
109-116
2019
10.1016/S0168-9002(03)01368-8
10.1667/RR3001.1
10.1016/j.bbcan.2015.06.008
10.1039/c2nr31227a
10.1667/RR1984.1
10.1158/0008-5472.CAN-07-6871
10.1088/1361-6560/aa54c9
10.1186/s12645-015-0012-3
2018
2018
Geant4 physics reference manual (2018) Geant4 physics reference manual. http://geant4-userdoc.web.cern.ch/geant4-userdoc/UsersGuides/PhysicsReferenceManual/fo/PhysicsReferenceManual.pdf [Accessed Nov 8, 2018]
10.1088/0031-9155/49/18/N03
10.1118/1.3476457
10.1016/j.ijrobp.2010.08.044
10.1016/j.jcp.2014.06.011
VF
Khokhlov
author
2006
2006
Nanoparticles in radiation therapy: a summary of various approaches to enhance radiosensitization in cancer.
D
Kwatra
author
2013
text
Translational Cancer Research
2013
2
4
330
342
S
Lehnert
author
2015
Radiosensitizers and Radiochemotherapy in the Treatment of Cancer.
2015
548 pp
10.1088/0031-9155/55/4/002
10.1038/srep00018
Dose-supplementary therapy of malignant tumors. Advances in Neutron Capture Therapy.
IN
Sheino
author
Y
Nakagawa
author
2006
text
Takamatsu
2006
531
534
IN
Sheino
author
2014
2014
10.1088/0031-9155/58/3/451
Advances in modern radiation therapy.
J
Van Dyk
author
J
Van Dyk
author
2005
text
Medical Physics Pub Corp
2005
1
31
10.3897/nucet.5.35798
https://nucet.pensoft.net/article/35798/
https://nucet.pensoft.net/article/35798/download/pdf/
https://nucet.pensoft.net/article/35798/download/xml/
It is experimentally proven that nanoparticles of high-Z materials can be used as radiosensitizers for photon beam therapy. In the authors’ opinion, data available as of today on the impact of secondary particles (electrons, photons and positrons generated in biological tissue by penetrating beam of primary photons) on the distribution of deposited dose during photon beam therapy in the presence of nanoparticles, are insufficient. Investigation of this impact constituted the main goal of this work.
Two-stage simulation was performed using Geant4 platform. During the first stage a layer of biological tissue (water) was irradiated by monoenergetic photon sources with energies ranging from 10 keV to 6 MeV. As the result of this modeling spectra of electrons, photons and positrons were obtained at the depth of 5 cm. During the second stage the obtained photon spectra were used to irradiate gold, gadolinium and water nanoparticles. Radial distributions of energy deposited around nanoparticles were obtained as the result of this modeling.
Radial DEF (Dose Enhancement Factor) values around nanoparticles of gold and gadolinium positioned in water at the depth of 5 cm were obtained after processing the collected data. Contributions from primary photons and secondary particles (electrons, photons and positrons generated in the layer of water with 5-cm thickness by the penetrating beam of primary photons) in the additional dose deposited around the nanoparticles were calculated as well.
It was demonstrated that layer of biological tissue placed between the source of photons and nanoparticles considerably changes the initial spectrum of photons and this change is significant in the analysis of mechanism of radiosensitization of biological tissues by nanoparticles for all energies of photon sources (up to 6 MeV).
It was established that interaction of electrons and positrons with nanoparticles does not lead to significant increase of additional dose in the vicinity of their surfaces and can be most likely excluded from consideration in the analysis of radiosensitization mechanism of nanoparticles.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Nanoparticles of gold and gadolinium
deposited dose
photon beam therapy
Geant4
Monte Carlo simulation
Impact of secondary particles on microdistribution of deposited dose in biological tissue in the presence of gold and gadolinium nanoparticles under photon beam irradiation
Research Article
10.3897/nucet.5.35800
2019-05-17
nucet
A.A. Bochvar High-technology Research Institute of Inorganic Materials, Moscow, Russia
author
Kuznetsov, Andrey
A.A. Bochvar High-technology Research Institute of Inorganic Materials, Moscow, Russia
author
Azovskov, Mikhail
A.A. Bochvar High-technology Research Institute of Inorganic Materials, Moscow, Russia
author
Belousov, Sergey
A.A. Bochvar High-technology Research Institute of Inorganic Materials, Moscow, Russia
author
Vereshchagin, Ilya
A.A. Bochvar High-technology Research Institute of Inorganic Materials, Moscow, Russia
author
Efremov, Alexey
A.A. Bochvar High-technology Research Institute of Inorganic Materials, Moscow, Russia
author
Khlebnikov, Sergey
2019-05-17
2019-05-17
2019
Nuclear Energy and Technology
2452-3038
5
117-122
2019
10.3897/nucet.5.35800
https://nucet.pensoft.net/article/35800/
https://nucet.pensoft.net/article/35800/download/pdf/
https://nucet.pensoft.net/article/35800/download/xml/
The article presents the results of work on dismantling the large installation equipment of Research Building B at the Bochvar High-technology Research Institute of Inorganic Materials (Bochvar Institute). The works were carried out as part of Building B preparation for decommissioning. The purpose of dismantling the large-sized capacitive equipment was to reconstruct the large installation site for managing radioactive waste generated during Building B decommissioning. The works on decommissioning a radioactively contaminated building within a densely populated district of megalopolis were carried out for the first time.
The characteristics of the large-sized capacitive equipment are presented. Radioactive contamination of the capacitive equipment is determined by long-lived a-emitting isotopes: 235U, 238U, 239Pu. The sequence of works on dismantling the radiation-contaminated capacitive equipment includes preparatory work, dismantling the tank piping, localizing radioactive contamination of the external surface of the equipment as well as dismantling and moving it into a transport container.
Dismantling and decontamination of the large-sized capacitive equipment was carried out by the Bochvar Institute Decommissioning Department. The following tools were used during the works: (1) a mobile foam decontamination facility to perform decontamination works and (2) a mobile high pressure facility to apply localizing and decontaminating film coatings. The tanks were dismantled by means of low-spark tools, i.e., reciprocating saws. Crane runways were made in order to move the dismantled equipment into transport containers: the movement was carried out with the help of a winch.
The main results of dismantling and decontaminating the radioactively contaminated tanks are the dismantling of four units of long-length column-type equipment with heights from 4.2 to 6.4 m and 26 units of capacitive equipment (maximum capacity = 8 m3) as well as decontamination of the internal surfaces of radiation-contaminated equipment (decontamination factor = 25–70). As a result, the activity of the accumulated radioactive waste was reduced (the RW class was changed from 3 to 4).
The main conclusion regarding the managment of large-sized radiation-contaminated tanks during Building B decommissioning is as follows: the works were organized and carried out at a high technical level, using modern decontamination and dismantling equipment and modern methods to ensure work safety at the Bochvar Institute site in the city of Moscow.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Decommissioning
radiation-contaminated equipment
tank
extraction column
dismantling
uranium
plutonium
foam decontamination
radioactive waste
Dismantling and decontamination of large-sized radiation-contaminated equipment during Research Building B decommissioning at the Bochvar Institute site
Research Article
10.3897/nucet.5.35801
2019-05-17
nucet
Saint Petersburg State Institute of Technology (Technical University), St. Petersburg, Russia
author
Chugunov, Aleksandr
Saint Petersburg State Institute of Technology (Technical University), St. Petersburg, Russia
author
Vinnitskii, Vadim
2019-05-17
2019-05-17
2019
Nuclear Energy and Technology
2452-3038
5
123-128
2019
10.3897/nucet.5.35801
https://nucet.pensoft.net/article/35801/
https://nucet.pensoft.net/article/35801/download/pdf/
https://nucet.pensoft.net/article/35801/download/xml/
Baromembrane purification methods as part of liquid radioactive media processing complexes are increasingly included in the practice of radioactive waste management. The paper presents the results of a comparative study of the performance of commercially available hyper- and nanofiltration elements when a simulated solution is continuously phosphatized. The study revealed the influence of changes in the feed solution salinity on the permeability, working pressure in the brine chamber of the hyper- and nanofiltration apparatus and the permeate salinity. It is shown that, in a closed loop of liquid radioactive waste, the introduction of polyphosphates to stabilize the truly dissolved forms of multivalent metals on the ULP reverse-osmotic membrane leads, as expected, to a systematic performance degradation, first of all, in the membrane permeability at a fixed pressure in the apparatus. The permeate of the system with a nanofiltration membrane, VNF (Vontron NanoFiltration), contains a sufficiently high salt concentration indicating that sodium salts formed during complexation are removed from the circuit, thereby reducing the solution osmotic pressure which critically affects the yield of the purified solution. Thus, nanofiltration in combination with chelating agents can be an effective tool for fractionating components of radioactive solutions, ensuring the achievement of standard indicators for wastewater and biologically hazardous substances that are subject to permanent disposal.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Radioactive waste
nanofiltration
complexation
osmotic pressure
reverse osmosis
membrane permeability
Nanofiltration fractionation of radioactive solution components as a method for reducing the volume of wastes intended for permanent disposal
Research Article
10.3897/nucet.5.36474
2019-06-21
nucet
National Research Nuclear University MEPhI, Obninsk, Russia
author
Starkov, Sergey
Kaluga Branch of the Bauman Moscow State Technical University, Kaluga, Russia
author
Lavrenkov, Yury
2019-06-21
2019-06-21
2019
Nuclear Energy and Technology
2452-3038
5
129-137
2019
10.3897/nucet.5.36474
https://nucet.pensoft.net/article/36474/
https://nucet.pensoft.net/article/36474/download/pdf/
https://nucet.pensoft.net/article/36474/download/xml/
Hydrogen energy is able to solve the problem of the dependence of modern industries on fossil fuels and significantly reduce the amount of harmful emissions. One of the ways to produce hydrogen is high-temperature water-steam electrolysis. Increasing the temperature of the steam involved in electrolysis makes the process more efficient. The key problem is the use of a reliable heat energy source capable of reaching high temperatures. High-temperature gas-cooled reactors with a gaseous coolant and a graphite moderator provide a solution to the problem of heating the electrolyte. Part of the heat energy is used for producing electrical energy required for electrolysis. Modern electrolyzers built as arrays of tubular or planar electrolytic cells with a nuclear energy source make it possible to produce hydrogen by decomposing water molecules, and the working temperature control leads to a decrease in the Nernst potential. The operation of such facilities is complicated by the need to determine the optimal parameters of the electrolysis cell, the steam flow rate, and the operating current density. To reduce the costs associated with the process optimization, it is proposed to use a low-temperature electrolysis system controlled by a spiking neural network. The results confirm the effectiveness of intelligent technologies that implement adaptive control of hybrid modeling processes in order to organize the most feasible hydrogen production in a specific process, the parameters of which can be modified depending on the specific use of the reactor thermal energy. In addition, the results of the study confirm the feasibility of using a combined functional structure made on the basis of spiking neurons to correct the parameters of the developed electrolytic system. The proposed simulation strategy can significantly reduce the consumption of computational resources in comparison with models based only on neural network prediction methods.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Spiking neural networks
high-temperature gas-cooled reactors
electro-optical neural commuting system
hydrogen production forecasting
centralized global parallel search circuit
Application of spiking neural networks for modelling the process of high-temperature hydrogen production in systems with gas-cooled reactors
Research Article
10.3897/nucet.5.36475
2019-06-21
nucet
Sevastopol State University, Sevastopol, Russia
author
Kachur, Svetlana
2019-06-21
2019-06-21
2019
Nuclear Energy and Technology
2452-3038
5
139-144
2019
10.3897/nucet.5.36475
https://nucet.pensoft.net/article/36475/
https://nucet.pensoft.net/article/36475/download/pdf/
https://nucet.pensoft.net/article/36475/download/xml/
The purpose of the study is to develop a model for predicting the process of a critical heat flux state with the VVER reactor core channel steaming. The model describes the dynamics of the nuclear reactor behavior in conditions of uncertainty, which are typical of abnormal situations, based on information on the process of heat exchange in the core process channels.
The use of the proposed model leads to an increase in the speed of response due to a simplified procedure to calculate the parameters of the heat exchange process in the reactor core. The quality of the reactor state assessment is improved through the prediction of the heat exchange process parameters and determination of the critical heat flux parameters in the core prior to the onset of surface boiling the potentiality of which is not predicted in modern VVER in-core monitoring systems.
A modification of the mathematical model has been proposed which offers the simplest possible way of using the advantages of neural networks in diagnostics. The model can be used to develop systems for diagnostics of in-core anomalies and systems for adaptive control of the VVER-type reactor thermal power.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Nuclear reactor
critical heat flux
power density
thermophysical model
identification
neural networks
Diagnostics of the critical heat flux state of a VVER reactor based on a channel steaming model
Research Article
10.3897/nucet.5.36476
2019-06-21
nucet
National Research Nuclear University MEPhI, Obninsk, Russia
author
Chusov, Igor
National Research Nuclear University MEPhI, Obninsk, Russia
author
Kirillov, Pavel
Private Institution «Atomstandart», Moscow, Russia
author
Pronyaev, Vladimir
ROSATOM State Atomic Energy Corporation, Moscow, Russia
author
Obysov, Nikolay
ROSATOM State Atomic Energy Corporation, Moscow, Russia
author
Novikov, Grigoriy
2019-06-21
2019-06-21
2019
Nuclear Energy and Technology
2452-3038
5
145-153
2019
10.3897/nucet.5.36476
https://nucet.pensoft.net/article/36476/
https://nucet.pensoft.net/article/36476/download/pdf/
https://nucet.pensoft.net/article/36476/download/xml/
The study is dedicated to the information technologies for storage, systematization and distribution of thermophysical data for nuclear power engineering. The general trend existing in the areas involving wide use of scientific data is the shifting from conventional databases to the development of a consolidated infrastructure capable of overcoming sharply growing volumes of scientific data with continuously increasing complexity of the data structure due to the expansion of the range of materials. The above infrastructure ensures interoperability, including data exchange and dissemination. The general principle of data management for thermophysical properties of the nuclear reactor materials based on the subject-oriented ReactorThermoOntology (RTO) is suggested in the present paper. The ontology includes a unified glossary of all concepts, expanded through logical connections and axioms. The suggested RTO ontology combines the terms typical for reactor materials, their characteristics, as well as all types of information entities determining textual, mathematical and computer structures. In the coded form, the ontology becomes the control add-in capable to integrate heterogeneous data. Its most important feature is the possibility of its permanent expansion, which is necessary with introduction of new materials and terms related to them, e.g. nanostructures characteristics. Beside the ontology, description of the reactor materials, the possible scenarios for the use of the ontology during the phases of design, operation and integration of autonomous resources, primarily databases, are examined in the paper. The use of Big Data technology with diverse variations of logical structures of the data is suggested as the most efficient tool for data integration. Joint use of the technologies which before were applied separately, such as exchange standard in the form of the structured text documents, data control based on the ontology and platform for the work with big data, allows the conversion of multiple existing primary resources (databases, files, archives, etc.) to the standard JSON text format for the subsequent semantic integration.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Thermophysical properties
reactor materials
nuclear fuel
ontology
database
data integration
JSON-format
Ontologies and databases on thermophysical properties of nuclear reactor materials
Research Article
10.3897/nucet.5.36477
2019-06-21
nucet
Saint-Petersburg State Institute of Technology (Technical University), St. Petersburg, Russia
author
Koryakovskiy, Yuriy
Saint-Petersburg State Institute of Technology (Technical University), St. Petersburg, Russia
author
Doilnitsyn, Valeriy
Saint-Petersburg State Institute of Technology (Technical University), St. Petersburg, Russia
author
Akatov, Andrey
https://orcid.org/0000-0002-1453-5837
2019-06-21
2019-06-21
2019
Nuclear Energy and Technology
2452-3038
5
155-161
2019
10.3897/nucet.5.36477
https://nucet.pensoft.net/article/36477/
https://nucet.pensoft.net/article/36477/download/pdf/
https://nucet.pensoft.net/article/36477/download/xml/
The article presents the results of work aimed at improving of chemical decontamination methods. A brief description of existing chemical decontamination technologies used to remove radioactive contamination (RC) from walls, floors and external surfaces of equipment without dismantling, i.e., in situ, is given. A vector of research aimed at improving the efficiency of fixed RC’s removal is also determined. The first aim of this work is to improve the decontaminating properties of removable polymer coatings used in practice. The following domestic products were chosen as study objects: compositions presented under trademarks VA, VL; and also special formulation ZPS-1M. Modifications of these compounds performed in SPSIT in some cases made it possible to significantly increase the decontamination factors (DF). The best results were obtained for VL compositions: it was found that due to certain additives it is possible to increase the DF for metal surfaces by a factor of 5–35 over the base product. Along with film-forming decontaminating compositions, an alternative patented technique has been developed in SPSIT. The main feature of this technique is usage of sorbent-based composite covering material previously saturated by decontaminating solution. New technique allows to achieve far higher decontamination factors (150–500) when fixed RC is removed from metal surfaces. In addition, it can be applied to polymer and other non-metal materials. One of the main advantages of given technique is a drastic (11–16 times) reduction of time required for carrying out decontamination operations. The obtained results may be useful for further research in this area. Thus, research performed allows to come up with general conclusion: there are possibilities to efficiently remove fixed RC from surfaces using rather simple chemical means. That, in its turn, could be a rational alternative to high-priced robotic decontaminating systems.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
decontamination
decommissioning
fixed radioactive contamination
film-forming compositions
composite decontamination material
Improving the efficiency of fixed radionuclides’ removal by chemical decontamination of surfaces in situ
Research Article
10.3897/nucet.5.36478
2019-06-21
nucet
JSC “SSC RF-IPPE named after A.I. Leypunsky”, Obninsk, Russia
author
Rykov, Nikita
JSC “SSC RF-IPPE named after A.I. Leypunsky”, Obninsk, Russia
author
Bezhunov, Gennady
JSC “SSC RF-IPPE named after A.I. Leypunsky”, Obninsk, Russia
author
Gorbachev, Vyacheslav
2019-06-21
2019-06-21
2019
Nuclear Energy and Technology
2452-3038
5
163-169
2019
10.3897/nucet.5.36478
https://nucet.pensoft.net/article/36478/
https://nucet.pensoft.net/article/36478/download/pdf/
https://nucet.pensoft.net/article/36478/download/xml/
The known dependence of absolute efficiency on energy and space for particular measurement conditions is used to determine the mass (activity) of 235U in solid radioactive waste by gamma-spectrometric method. The ISOCS system makes it possible to avoid laborious and time-consuming calibration measurements using standard samples to obtain the absolute efficiency curve due to using the so-called characterized detector having a file with a set of efficiencies for various measurement geometries.
In many cases, the establishment of standard samples with parameters covering the 235U mass measurement range in the variation intervals of influencing factors, including density, non-uniformity, isotopic composition, geometry, etc., is very expensive and, most often, not feasible. With regard for this, a computational and experimental approach is used based on results obtained by Monte Carlo method using the MCNP code with variation of the key influencing parameters in a broad range.
Calculations were performed for detector-recorded spectra of gamma quanta from casks containing waste differing in the density of the cask content (the density was calculated with regard for the uranium contained in waste) – from 0.016 to 1 g/cm3, in the mass of uranium in waste – from 0.64 g to 2 kg, and in the matrix material – graphite, cellulose, quartz, cellulose with 20 % of iron dust.
Applicability boundaries have been defined for the developed procedure to measure uranium-containing waste in terms of the material matrix (~ 2.2 %) and its density (~ 10 %) and the contribution of the uranium mass uncertainty in the cask (5 % for nonporous matrices, 10 % for porous matrices) to the obtained result has been estimated.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Nondestructive analysis of nuclear materials
solid radioactive waste
uranium mass
gamma-spectrometry
ISOCS system
absolute efficiency curve
Monte Carlo method
MCNP code
measurement procedure
measurement procedure range
Use of mathematical modeling to extend the scope of application for the procedure of measuring the mass of 235U in solid radioactive waste
Research Article
10.3897/nucet.5.36479
2019-06-21
nucet
Sosny R&D Company, Dimitrovgrad, Russia
author
Kuzmin, Ilya
Sosny R&D Company, Moscow, Russia
author
Leshchenko, Anton
DETI MEPhI, Dimitrovgrad, Russia
Sosny R&D Company, Dimitrovgrad, Russia
author
Pavlov, Sergey
Sosny R&D Company, Dimitrovgrad, Russia
DETI MEPhI, Dimitrovgrad, Russia
author
Shamsutdinov, Rinat
Innovation and Technology Center for the Proryv Project, Moscow, Russia
author
Mochalov, Yuriy
2019-06-21
2019-06-21
2019
Nuclear Energy and Technology
2452-3038
5
81-85
2019
10.3897/nucet.5.36479
https://nucet.pensoft.net/article/36479/
https://nucet.pensoft.net/article/36479/download/pdf/
https://nucet.pensoft.net/article/36479/download/xml/
Nuclear fuel pellets are sintered in high-temperature furnaces in an atmosphere with strictly defined requirements for the composition of the gas environments in the furnace’s different temperature zones. The preset process conditions in the mixed nitride uranium-plutonium (MNUP) fuel pellet sintering furnace is achieved through the respective gas supply arrangement and by the design of the barriers between the temperature zones and that of the gas supply and discharge units. A CFD model was created in the Ansys Fluent package and validated for testing the functionality of the design concepts used to develop the MNUP fuel sintering furnace channel. A mockup of the sintering furnace channel, which makes a part of the gas-dynamic test bench, was developed and fabricated for the analytical model validation.
The paper presents a description of the test bench design and performance for measuring the concentration of gases in the channel simulating the nitride nuclear fuel sintering furnace channel. The results of the test bench gas-dynamic studies were used for the computational and experimental justification of the approaches used to develop the sintering furnace channel. The functionality of the barriers for the sintering furnace channel division into zones with the preset composition of the gas environments and the gas supply and discharge units has been tested experimentally. The obtained experimental data on the distribution of the process gas concentration makes it possible to validate computational thermophysical and gas-dynamic CFD models of the MNUP fuel sintering furnace channel.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
gas-dynamic studies
sintering furnace
furnace channel
furnace zone
MNUP fuel
gas distribution modeling
gas concentration measurement
gas sampling
barrier
Test bench for gas-dynamic studies in the furnace channel for nuclear fuel pellet sintering
Research Article
10.3897/nucet.5.34590
2019-06-21
nucet
University of Kyrenia, Girne,
author
Abbasi, Akbar
Near East University, Lefkosa,
author
Sadikoglu, Fahreddin
2019-06-21
2019-06-21
2019
Nuclear Energy and Technology
2452-3038
5
177-182
2019
10.1016/j.nucengdes.2018.01.005
10.1016/j.ress.2016.07.019
10.1016/j.anucene.2015.10.033
10.1016/j.nucengdes.2003.11.022
10.1016/j.pnucene.2009.09.004
10.1109/ECAI.2015.7301174
10.1016/j.aei.2005.01.009
10.1007/978-3-7908-1852-9_1
10.1016/j.pnucene.2005.03.002
10.1016/j.anucene.2014.10.036
10.1016/S0967-0661(97)00046-4
10.1016/j.anucene.2015.04.028
10.1063/1.5025993
10.1109/ICEETS.2013.6533515
D
Ruan
author
2013
2013
10.1109/ICENCO.2015.7416342
10.13182/NT10-A9451
10.1016/j.applthermaleng.2017.07.045
10.1016/j.anucene.2005.12.008
10.3897/nucet.5.34590
https://nucet.pensoft.net/article/34590/
https://nucet.pensoft.net/article/34590/download/pdf/
https://nucet.pensoft.net/article/34590/download/xml/
Nowadays, Nuclear Power Plant (NPP) is one of the intended energy resources for the world requirement energy in future, and nuclear power plants provided 11 percent of the world’s electricity production in 2014. Meanwhile, nuclear power plant safety has always been one of the most critical issues in the world. In this paper, the nuclear power plant safety improvement using Soft Computing Techniques were analyzed. For this purpose, the support system based on Neuro-Fuzzy Diagnosis System (NFDs) method and Genetic Algorithms (GAs) approach were used. The obtained result showed that the first symptom is P3 (pressurizer pressure) and second order symptom is P2 (core coolant average temperature) in both approaches. The comparison between the NFDs method and the GAs approaches indicated that the GAs in data test results was faster than the NFDs results.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Nuclear Power Plant (NPP)
Soft Computing
Neuro Fuzzy Diagnosis System (NFDs)
Genetic Algorithms (GAs)
Safety
Application of soft computing techniques in WWER nuclear power plant safety
Research Article
10.3897/nucet.5.38117
2019-09-25
nucet
University of Pisa, Pisa, Italy
author
D’Auria, Francesco
https://orcid.org/0000-0001-7638-7801
University of Zagreb, Zagreb, Croatia
author
Debrecin, Nenad
GRS, Munich, Germany
author
Glaeser, Horst
2019-09-25
2019-09-25
2019
Nuclear Energy and Technology
2452-3038
5
183-199
2019
10.3897/nucet.5.38117
https://nucet.pensoft.net/article/38117/
https://nucet.pensoft.net/article/38117/download/pdf/
https://nucet.pensoft.net/article/38117/download/xml/
The present paper deals with the proposal of an additional safety barrier for the class of large (1000 MWe or more) Light Water Reactors (LWR) now in operation, in construction, or under design. Emphasis is given to the motivations or the needs for the barrier. Two main parts of the paper can be distinguished. The following topics are discussed in the former part (section 2): (a) the weakness of the barrier constituted by the current design of nuclear fuel; (b) the continuously increasing complexity of the system, with main reference to the Instrumentation and Control (I&C); (c) the role that the Large Break Loss of Coolant Accident (LBLOCA) had for arriving at the current layout of the Reactor Coolant System (RCS). Furthermore avoiding the severe accidents in 1979, 1987 and 2011, is at the basis of the proposal. In the latter part (sections 3 and 4), the elements of the proposed technological safety barrier are discussed: the As-Low-As-Reasonably-Achievable (ALARA) principle, the Best Estimate Plus Uncertainty (BEPU) approach, the Extended Safety Margin Detection (E-SMD) hardware, the Emergency Rescue Team (ERT) strategy (or a virtual entity for the reactor) and the Independent Assessment (IA) concept. The additional safety barrier, although not demonstrated in the paper, is expected to reduce for a factor in the range 10–1000 the probability of core melt and to have a cost in the order of 1% the cost of a nuclear reactor unit.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Nuclear reactor safety
BEPU
licensing
extended safety margins
additional safety barrier
The technological challenge for current generation nuclear reactors
Research Article
10.3897/nucet.5.39239
2019-09-25
nucet
National Research Nuclear University MEPhI, Moscow, Russia
author
Putilov, Alexandr
National Research Nuclear University MEPhI, Moscow, Russia
author
Strikhanov, Mikhail
https://orcid.org/0000-0003-2586-0405
National Research Nuclear University MEPhI, Moscow, Russia
author
Tikhomirov, Georgy
https://orcid.org/0000-0002-5332-7272
2019-09-25
2019-09-25
2019
Nuclear Energy and Technology
2452-3038
5
201-206
2019
NA
Ilyina
author
2012
2012
Digital Future: The next step in the development of nuclear energy technologies.
VV
Ivanov
author
2017
text
Energeticheskaya Politika [Energy Policy]
2017
3
31
42
Alexandr Illich Leipunsky and his principles in the system of higher Education. Izvestia vuzov. Yadernaya Energetika [News of Higher Educational Institutions.
PL
Kirillov
author
2018
text
Nuclear Energy]
2018
1
165
168
AV
Putilov
author
2010
Innovation activity in the nuclear industry. Book one. “Basic principles of innovation policy”.
2010
184 pp
MN
Strikhanov
author
2006
Science in Russia. Sociological analysis.
2006
455 pp
MN
Strikhanov
author
2007
Higher school of Russia from the position of nonlinear dynamics (problems, estimations, models).
2007
192 pp
10.3897/nucet.5.39239
https://nucet.pensoft.net/article/39239/
https://nucet.pensoft.net/article/39239/download/pdf/
https://nucet.pensoft.net/article/39239/download/xml/
The article briefly describes the history of training personnel for the nuclear industry and sets tasks for its improvement and development to ensure the future growth of this industry. Within the framework of the emerging digital economy, such a phenomenon as digital platforms erases the boundaries between industries, forming new unexpected industrial alliances, even new industries. Innovative activities in the power industry, including nuclear power, should provide the possibility of forming digital economic platforms in various energy segments as well as training personnel in using this new toolkit. Today, the Rosatom State Nuclear Energy Corporation is developing more than 30 projects of new nuclear power plants (NPP) in Russia and 12 other countries. This requires educational support, and for this purpose a Consortium of supporting universities of the Rosatom State Corporation was established, which includes 18 specialized higher educational institutions. More than half of them train personnel directly for designing, constructing and operating NPPs. The scale of the necessary personnel training in the near future indicates that we need a new “educational paradigm”, which can be described as “front-line education”, i.e., training personnel for developing digital economy technologies simultaneously along the entire “front”. This “front” stretches from schoolchildren preparing to enter universities to production personnel whose professional development should be carried out taking into account the specifics of the digital transformation of production. Partnership is one of the leading values of the modern young generation. To withstand high competition for the best personnel, organizations must not only be saturated with the culture of partnership from the inside but also act as reliable partners for one another in involving and training young employees.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Nuclear power
educational technology
digital economy in the nuclear industry
personnel development
export of education
Personnel training for the developing nuclear power industry
Research Article
10.3897/nucet.5.39288
2019-09-25
nucet
State Scientific Center – Research Institute of Atomic Reactors, Dimitrovgrad, Russia
author
Alekseev, Aleksandr
State Scientific Center – Research Institute of Atomic Reactors, Dimitrovgrad, Russia
author
Dreganov, Oleg
State Scientific Center – Research Institute of Atomic Reactors, Dimitrovgrad, Russia
author
Izhutov, Aleksey
State Scientific Center – Research Institute of Atomic Reactors, Dimitrovgrad, Russia
author
Kiselyova, Irina
State Scientific Center – Research Institute of Atomic Reactors, Dimitrovgrad, Russia
author
Shulimov, Vitalij
2019-09-25
2019-09-25
2019
Nuclear Energy and Technology
2452-3038
5
207-212
2019
Test methods in the MIR reactor of VVER fuel during transient and emergency conditions. News of Higher Educational Institutions.
AV
Alekseev
author
2007
text
Nuclear Power Engineering
2007
3
1
83
91
10.1007/s10512-012-9614-6
Examination of VVER fuel rod behavior under RIA. In-reactor experiment method and process.
AV
Alekseev
author
2006a
text
RIAR collected papers
2006a
1
23
32
10.1007/s10512-006-0185-2
Post-test processing method and results for the data generated when testing VVER-1000 fuel in the MIR reactor under RIA.
AV
Alekseev
author
2008
text
RIAR collected papers
2008
4
66
70
10.1007/s10512-011-9337-0
Use of the MUZA program for computational support of experiments in research reactors.
VV
Alekseev
author
2013
text
Nuclear Reactor Physics
2013
3
135
140
10.1007/BF02673204
On the Accuracy of Describing Different Codes of Critical Heat Flows in Rods Bundles.
VP
Bobkov
author
2001
text
Heat Power Engineering
2001
3
21
28
10.1134/S0040601511040045
VV
Bolshakov
author
2009
2009
VV
Bolshakov
author
2009a
2009a
AV
Burukin
author
2009
2009
Results of the “Steady-State Crisis” Experiment.
OI
Dreganov
author
2014
text
RIAR Collected Papers
2014
2
3
9
OI
Dreganov
author
2015
2015
Integral reactor experiments to test multi-component fragments of VVER-440 and VVER-1000 FAs under LOCA conditions. Summary of experimental results.
AV
Goryachev
author
2004
text
Nuclear Reactor Physics
2004
2
29
38
10.1016/j.nucengdes.2007.02.014
AL
Izhutov
author
2015
2015
DA
Krylov
author
1995
1995
SA
Logvinov
author
2004
2004
VV
Lozhkin
author
1998
1998
O
Nechayeva
author
2003, 2004
2003, 2004
VV
Sergeev
author
1998
1998
VI
Shchekoldin
author
1998
1998
10.3897/nucet.5.39288
https://nucet.pensoft.net/article/39288/
https://nucet.pensoft.net/article/39288/download/pdf/
https://nucet.pensoft.net/article/39288/download/xml/
To license nuclear fuel for nuclear power plants, data on the behavior of fuel elements (FE) under design-basis accidents are required. These data are obtained during tests of fuel assemblies (FA) and single fuel elements in research reactor channels followed by post-test studies in protective chambers.
A reactivity-initiated accident (RIA) with an unauthorized release of CPS rods from the reactor core leads to a pulsed channel power increase. This accident can proceed according to two scenarios: without a critical heat flux (CHF) on the fuel element jacket at the final stage and with a dry heat flux. To date, a series of experiments have been carried out according to the first scenario in the MIR reactor channel and the corresponding data on the behavior of fuel elements have been obtained. An urgent task for today is to prepare and conduct reactor experiments according to the second scenario.
The main experimental parameter that determines the behavior and final state of the studied fuel elements is their temperature. No experimental data were found on the critical heat flux for the rod bundles in the low coolant mass flow rate region (experiments in the MIR reactor channel can be conducted in the range of 200–250 kg/(m2s)). The available data are in the extrapolation range.
The “steady-state crisis” experiment was conducted to obtain data on the critical heat flux value within the specified coolant mass flow rate range in the MIR reactor channel. The test object was a jacket fuel assembly composed of three shortened VVER-1000 fuel rods with a length of 1230 mm (the fuel part length = 1000 mm) installed in a triangular grid at a pitch of 12.75 mm, which is a cell of the VVER-1000 core. This assembly configuration is used for in-pile tests to study the behavior of fuel elements under emergency conditions.
The in-pile testing results are presented. The paper shows the possibility of detecting the start and development of a dry heat flux based on the readings of thermocouples located inside the FE kernel. As a result, the directly measured test parameters were used to determine the critical heat flux value.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
MIR reactor
fuel element
experimental fuel assembly (EFA)
critical heat flux (CHF)
RIA (reactivity-initiated accident)
thermocouple (TC)
temperature
coolant flow rate
Outcomes of the “steady-state crisis” experiment in the MIR reactor channel
Research Article
10.3897/nucet.5.39290
2019-09-25
nucet
JSC RPA «CNIITMASH», Moscow, Russia
author
Orlov, Viktor
JSC RPA «CNIITMASH», Moscow, Russia
author
Anosov, Nikolay
JSC RPA «CNIITMASH», Moscow, Russia
author
Skorobogatykh, Vladimir
JSC RPA «CNIITMASH», Moscow, Russia
author
Gordyuk, Lyubov
JSC RPA «CNIITMASH», Moscow, Russia
author
Yurgina, Zhanna
JSC Institute of Reactor Materials, Zarechny, Russia
author
Koshcheev, Konstantin
JSC Institute of Reactor Materials, Zarechny, Russia
author
Barsanova, Svetlana
JSC «SSC RIAR», Dimitrovgrad, Russia
author
Shamardin, Valentin
2019-09-25
2019-09-25
2019
Nuclear Energy and Technology
2452-3038
5
213-218
2019
10.3897/nucet.5.39290
https://nucet.pensoft.net/article/39290/
https://nucet.pensoft.net/article/39290/download/pdf/
https://nucet.pensoft.net/article/39290/download/xml/
Brittle fracture resistance of RPV 15H2NMFA grade 1 steel is investigated. Sets of small-sized testing samples and a set of standard-sizes samples were used in the study. It was demonstrated that application of sets of small-sized specimens in mechanical tests for determining the brittle fracture resistance of RPV 15H2NMFA grade 1 steel makes possible the following:
increasing the volume of tests in each batch by 8 times without significant changes in the design of irradiation facility thus ensuring maintaining the initial irradiation parameters during testing;
substantially expanding the database of test results for statistical processing.
The need for large-scale modeling of process conditions arising in weld joint zones inaccessible for direct testing, such as: (1) the welding zone between the base metal and the corrosion-resistant coating metal, (2) the welding area between the weld metal and the corrosion-resistant coating metal, and (3) the fusion area between the base metal, the weld metal, and the anticorrosive cladding metal, is demonstrated in the paper.
Process modeling of the metal in fusion areas up to 0.5 mm wide (each is 100 μm in size) with an experimental electroslag refined (ESR) ingot of up to 300 mm long with similar microstructure and variable chemical composition allows the following: (1) examining not less than 1000 small-sized impact testing samples with continuous distribution of concentrations of chemical elements in accordance with a certain law; and (2) testing these samples and identifying brittle fracture dangerous zones across fusion areas between the base metal and the anti-corrosive padding metal in the initial state or after subsequent irradiation at a given fluence rate and temperature.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Assessment of brittle fracture resistance
critical brittle point TK
ductile-to-brittle-transition temperature TP
conservative vessel life estimations
padding and welded joints
The study of the brittle fracture resistance in fusion areas between RPV steel 15H2NMFA grade 1 and austenitic padding
Research Article
10.3897/nucet.5.39318
2019-09-25
nucet
Ivanovo State Power University, Ivanovo, Russia
author
Gorbunov, Vladimir
Ivanovo State Power University, Ivanovo, Russia
author
Ivanova, Natalya
Ivanovo State Power University, Ivanovo, Russia
author
Lonshakov, Nikita
Ivanovo State Power University, Ivanovo, Russia
author
Belov, Yaroslav
2019-09-25
2019-09-25
2019
Nuclear Energy and Technology
2452-3038
5
219-224
2019
10.1299/jsmeicone.2015.23._ICONE23-1_87
VM
Batenin
author
2017
2017
BA
Dementyev
author
1990
1990
A
Dolgov
author
2016
2016
Influence of the neutron flux depression in the RBMK cell on the magnitude of the maximum and average fuel temperature.
AO
Goltsev
author
2009
text
Izvestiya Tomskogo Politekhnicheskogo Universiteta
2009
314
4
5
7
Experience in using the software complex at Ivanovo State Power Engineering University named after V.I. Lenin.
VA
Gorbunov
author
2011a
text
ANSYS Advantage (Russkaya Redaktsiya)
2011a
15
38
39
Predicting the accuracy of results in solving heat transfer problems based on neural network technologies.
VA
Gorbunov
author
2011b
text
Promyshlennaya Energetika,
2011b
8
48
52
Calculations for the optimization of geometrical and operating parameters of VVER-SKD fuel assemblies for different modes of the reactor operation with supercritical water parameters. Izvestiya Vysshykh Uchebnykh Zavedeniy.
KV
Kartashov
author
2012
text
Yadernaya Energetika
2012
2
3
11
GN
Kolpakov
author
2009
Designs of Fuel Elements, Channels and Cores of Power Reactors.
2009
118 pp
ST
Leskin
author
2011
Physical Features and Design of the VVER-1000 Reactor.
2011
116 pp
VS
Loginov
author
2009
Approximate Methods of Thermal Calculation for Active Elements of Electrophysical Plants.
2009
273 pp
2018
2018
New Generation Codes (2018) New Generation Codes. http://www.ibrae.ac.ru/contents/68/ [accessed Jun 10, 2018] [in Russian]
LD
Palmer
author
1961
1961
Modeling of the thermal-engineering reliability of a fuel rod with different options of the power density and temperature variation.
RR
Perimov
author
2004
text
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2004
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150
153
TS
Shcherbakova
author
2008
2008
Procedures and program of the stationary temperature field calculation in the system of multi-zone cylindrical fuel rods. Izvestiya Vysshykh Uchebnykh Zavedeniy.
VA
Starkov
author
2013
text
Yadernaya Energetika
2013
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54
62
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PA
Ushakov
author
1972
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Teplofizika Vysokikh Temperatur
1972
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Velesyuk
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AD
Vishnyakova
author
2015
text
Yadernaya Energetika
2015
4
61
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Systematization of the studies on heat exchange in fuel assemblies and selected problems of liquid metal cooling. Izvestiya Vysshykh Uchebnykh Zavedeniy.
AV
Zhukov
author
2009
text
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10.3897/nucet.5.39318
https://nucet.pensoft.net/article/39318/
https://nucet.pensoft.net/article/39318/download/pdf/
https://nucet.pensoft.net/article/39318/download/xml/
Water-cooled water-moderated reactors (VVER) are widely used at Russian nuclear power plants. The VVER reactor core is formed by fuel assemblies consisting of fuel rods. The fuel in fuel rods is uranium dioxide. The safety of the reactor operation is ensured through stringent requirements for the maximum nuclear fuel temperature. Calculation of temperature fields within the reactor core requires associated problems to be solved to determine the internal energy release in fuel based on neutronic characteristics. Dedicated software for such calculations is not available to a broad range of users. At the present time, there are numerical thermophysical modeling packages available for training or noncommercial applications which are used extensively, including Elcut, Flow Vision, Ansys Fluent, and Comsol Multiphysics. Verification of the obtained results is becoming an important issue in building models using these calculation packages.
An analytical solution was obtained as part of the study for the fuel temperature field determination. A program was developed in MathCAD based on this solution. A model was developed in Comsol Multiphysics to determine the fuel temperature field with constant thermophysical properties in a two-dimensional problem statement. The numerical model was verified using the analytical solution. The influence of the number of the grid nodes on the solution accuracy was established. The analytical solution can be used to determine the fuel temperature field at any radial coordinate of the reactor. The temperature field determination model developed in MathCAD can be used to verify numerical models of the fuel temperature field determination developed in dedicated packages.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Fuel element
fuel temperature field
analytical solution
mathematical model to determine fuel temperature fields
verification of numerical calculations
influence of tuning coefficients
safety of fuel element heating
Development of the model to determine the fuel temperature field in a two-dimensional problem statement
Research Article
10.3897/nucet.5.39319
2019-09-25
nucet
Nizhny Novgorod State Technical University n.a. R.Ye. Alekseev, Nizhny Novgorod, Russia
author
Dmitriev, Sergey
Nizhny Novgorod State Technical University n.a. R.Ye. Alekseev, Nizhny Novgorod, Russia
author
Mamaev, Alexandr
Nizhny Novgorod State Technical University n.a. R.Ye. Alekseev, Nizhny Novgorod, Russia
author
Rayzapov, Renat
Nizhny Novgorod State Technical University n.a. R.Ye. Alekseev, Nizhny Novgorod, Russia
author
Sobornov, Aleksey
Nizhny Novgorod State Technical University n.a. R.Ye. Alekseev, Nizhny Novgorod, Russia
author
Kotin, Andrey
JSC Afrikantov OKBM, Nizhny Novgorod, Russia
author
Bescherov, Dmitry
JSC Afrikantov OKBM, Nizhny Novgorod, Russia
author
Bolshuhin, Mikhail
2019-09-25
2019-09-25
2019
Nuclear Energy and Technology
2452-3038
5
225-229
2019
10.3897/nucet.5.39319
https://nucet.pensoft.net/article/39319/
https://nucet.pensoft.net/article/39319/download/pdf/
https://nucet.pensoft.net/article/39319/download/xml/
One of the most important scientific and technical tasks of the nuclear power industry is to assure the reactor equipment life and reliability under random temperature pulsations. High-intensity temperature pulsations appear during the process of mixing non-isothermal coolant flows. Coolant thermal pulsations cause corresponding, sometimes very significant, fluctuations in the temperature stresses of the heat-exchange surface metal, which, added to static loads, can lead to fatigue failure of equipment components.
The purpose of this work was to conduct an experimental study of the temperature and stress-strain states of a pipe sample under the influence of local stochastic thermal pulsations caused by the mixed single-phase heat coolant flows.
To solve the set problems, an experimental section was created, which made it possible to simulate the process of mixing non-isothermal coolant flows accompanied by significant temperature pulsations. The design of the experimental section allowed us to study the thermohydraulic and life characteristics of pipe samples made of austenite steel (60×5 mm). Some tools were developed for measuring the pipe sample stress-strain state and the coolant flow temperature field in the zone of mixed single-phase media with different temperatures. The measuring tools were equipped with microthermocouples and strain sensors.
As a result, we obtained experimental data on temperature pulsations, time-averaged temperature profiles of the coolant flow in the mixing zone as well as statistical and spectral-correlation characteristics of thermal pulsations. Based on the results of measuring the relative strains, the values of fatigue stresses in the mixing zone were calculated.
In addition, some devices and methods were elaborated to measure the temperature and stress-strain states of the pipe sample under the influence of local stochastic thermal pulsations. The developed experimental section provided thermal-stress loading of the metal surface at a high level of alternating stress amplitudes causing rapid damage accumulation rates. The results were included in the database to verify the method for assessing the fatigue life of structural materials for nuclear power plants as applied to austenite steel 12Cr18Ni10Ti under the influence of random thermal cyclic loads.
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en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Equipment life
temperature pulsations
coolant
temperature field
stress-strain state
Experimental studies of temperature pulsations during the process of mixing non-isothermal coolant flows in nuclear reactor equipment components
Research Article
10.3897/nucet.5.39320
2019-09-25
nucet
Obninsk Institute for Nuclear Power Engineering, Obninsk, Russia
author
Godes, Alexander
National Research Nuclear University MEPhI, Moscow, Russia
author
Kudriavtseva, Anna
Obninsk Institute for Nuclear Power Engineering, Obninsk, Russia
author
Shablov, Vladimir
2019-09-25
2019-09-25
2019
Nuclear Energy and Technology
2452-3038
5
231-235
2019
10.3897/nucet.5.39320
https://nucet.pensoft.net/article/39320/
https://nucet.pensoft.net/article/39320/download/pdf/
https://nucet.pensoft.net/article/39320/download/xml/
The purpose of the present paper is the formulation of the analytical version of the resonance coupled-channel model (RCCM) originally developed for D + T → 5He** → α + n nuclear fusion reaction. The integral in the denominator of the Breit-Wigner type is examined in the expression for S-matrix elements of binary processes in this model. Imaginary part of this integral determines the energy-dependent decay width for the near-threshold channel. It is demonstrated that this integral can be calculated explicitly with the Binet representation for the ψ-function (the logarithmic derivation of the gamma function). As the result the explicit expression for the S-matrix elements in the form of analytical functions of the channel momenta are obtained and the equivalence of the RCCM and the effective range approximation (Landau – Smorodinsky – Bethe approximation) is established on this basis. This allows expressing the parameters of the RCCM through the model independent system characteristics: the complex scattering length and the complex effective range. Several sets of model parameters of both approaches that provide a good description of the measured data on D + T → α + n reaction and D-T elastic scattering are derived. By this means we find the location of the S – matrix poles on different Riemann sheets which corresponds to Jπ = (3/2)+ state of 5He and 5Li nuclei. In particular, the location of the resonance (R) and shadow (S) poles is determined:
5He**: ZR = 46.9 – i37.2 (keV) ZS = 81.7 – i3.5 (keV)
5Li**: ZR = 205.7 – i146.8 (keV) ZS = 264.4 + i112.0 (keV).
Our results agree well with previous findings. The possible generalizations of the results obtained are discussed.
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en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Thermonuclear reactions
resonance coupled-channel model
effective range approximation
S-matrix poles
resonance and shadow poles
Jπ = (3/2)+ state of 5He and 5Li nuclei
Analytical version of the resonance coupled-channel model for D + T → 5He** → α + n reaction and its application for the description of low-energy D-T and D- 3He scattering
Research Article
10.3897/nucet.5.39340
2019-09-25
nucet
Nizhny Novgorod State Technical University, Nizhny Novgorod, Russia
author
Beznosov, Aleksander
Nizhny Novgorod State Technical University, Nizhny Novgorod, Russia
author
Lvov, Aleksander
Nizhny Novgorod State Technical University, Nizhny Novgorod, Russia
author
Bokov, Pavel
Nizhny Novgorod State Technical University, Nizhny Novgorod, Russia
author
Bokova, Tatyana
Nizhny Novgorod State Technical University, Nizhny Novgorod, Russia
author
Lukichev, Nikita
2019-09-25
2019-09-25
2019
Nuclear Energy and Technology
2452-3038
5
237-240
2019
Specific Features of Lead and Lead Bismuth Cooled Reactor Circuit Circulation Pumps. Izvestiya Vysshykh Uchebnykh Zavedeniy.
AV
Beznosov
author
2009
text
Yadernaya Energetika
2009
4
155
160
AV
Beznosov
author
2016
Technologies and Main Components of Lead and Lead Cooled Reactor Plant and Commercial and Test Facility Circuits.
2016
488 pp
Hydrodynamics and efficiency of the BREST-OD-300 RCP models in lead coolant at the FT-4 NGTU test facility. Proc. of the NNSTU n.a. R.Ye.
AV
Beznosov
author
2015a
text
Alekseyev
2015a
1
125
133
AV
Beznosov
author
2014
2014
AV
Beznosov
author
2012
Components of Power Circuits with Heavy Liquid Metal Coolants in Nuclear Power.
2012
536 pp
Experimental research and development of lead and lead-bismuth coolant pumps for nuclear plants. Proc. of the NNSTU n.a. R.Ye.
AV
Beznosov
author
2017a
text
Alekseyev
2017a
1
117
128
AV
Beznosov
author
2006
Heavy Liquid Metal Coolants in Nuclear Power.
2006
435 pp
Features of a high-temperature lead cooled axial pump wet end performance. Proc. of the NNSTU n.a. R.Ye.
AV
Beznosov
author
2013
text
Alekseyev
2013
4
101
206
213
10.3103/S1068366614040035
Experimental study of the wet end of a heavy liquid metal cooled reactor axial coolant pump model.
AV
Beznosov
author
2014b
text
Vestnik Mashinostroyeniya
2014b
2
38
45
Experimental studies into the dependencies of the axial lead coolant pump performance on the parameters of the impeller profile grids. Izvestiya Vuzov.
AV
Beznosov
author
2017b
text
Yadernaya Energetika
2017b
1
138
144
Experimental research and development of the lead coolant pump characteristics. Izvestiya Vysshykh Uchebnykh Zavedeniy.
AV
Beznosov
author
2015b
text
Yadernaya Energetika
2015b
4
123
132
AV
Beznosov
author
2012
2012
VM
Budov
author
1986
Pumps of Nuclear Power Plants.
1986
408 pp
AV
Chechyotkin
author
1971
High-Temperature Coolants.
1971
496 pp
BREST fast neutron lead cooled reactor.
YuG
Dragunov
author
2015
text
Problemy Mashinostroyeniya i Avtomatizatsii
2015
3
97
103
YuN
Drozdov
author
2009
2009
1987
Vol. 1. Ed. B.S. Petukhov.
1987
360 pp
Handbook on Thermohydraulic Analysis (1987) Vol. 1. Ed. B.S. Petukhov.Energoatomizdat Publ., Moscow, 360 pp. [in Russian]
VYa
Karelin
author
1975
Cavitation Phenomena in Centrifugal and Axial Pumps.
1975
336 pp
AA
Lomakin
author
1966
Centrifugal and Axial Pumps.
1966
364 pp
AK
Mikhaylov
author
1977
Paddle Pumps. Theory, Calculation and Design.
1977
288 pp
K
Pfleiderer
author
1960
Paddle Machines for Liquid and Gas.
1960
685 pp
VV
Rozhdestvenskiy
author
1977
1977
10.3897/nucet.5.39340
https://nucet.pensoft.net/article/39340/
https://nucet.pensoft.net/article/39340/download/pdf/
https://nucet.pensoft.net/article/39340/download/xml/
The paper presents the results of experimental studies into the dependences of the axial pump performance (delivery rate, head, efficiency) in lead coolant on the parameters of the straightening device (SD) installed downstream of the impeller (the SD inlet flow angle and the number of the SD blades with a variable impeller speed change).
The studies were performed as applied to the operating conditions of small and medium plants with lead cooled fast neutron reactors with horizontal steam generators (BRS GPG). The designs of such plants are being matured at Nizhny Novgorod State Technical University (NNSTU).
The experiments were conducted on the FT-4 NGTU test bench at the lead coolant temperatures in a range of 440 to 500 °C. The number of the test blades was five and eight, and the SD inlet flow angle was 22, 24, 28, and 32°. The tests were also performed without an SD (with the SD dismantled). The shaft speed of the NSO-01 NGTU pump, with changeable SDs installed into its rotating assembly, was varied in a range of 600 to 1100 rev/min with a step of 100 rev/min. The SD sleeve diameter was 82 mm, the SD blade diameter and height were 213 mm and 80 mm respectively, and the maximum lead coolant flow rate during the studies was up to ~ 1650 t/h. The NSO-01 NGTU pump performance was determined with four changeable straightening devices and with no SD, the pump shaft speed being 600 to 1100 rev/min, as the circulation circuit hydraulic resistance changed owing to the movement of the wedge in the valve installed in it. The tests were performed with the impeller designed and supplied by NNSTU (D = 213 mm, dsl = 82 mm, the blade number is four, and the blade angle is 28°).
The obtained results are recommended for use to design axial heavy liquid metal coolant pumps.
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en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Heavy liquid metal coolant
fast neutron reactor
lead coolant
axial pump
pump head
pump delivery rate
pump impeller
Experimental studies into the dependences of the axial lead coolant circulation pump performance on the pump straightening device parameters
Research Article
10.3897/nucet.5.46379
2019-09-25
nucet
National Research Nuclear University MEPhI, Moscow, Russia
author
Murogov, Victor
2019-09-25
2019-09-25
2019
Nuclear Energy and Technology
2452-3038
5
241-248
2019
10.3897/nucet.5.46379
https://nucet.pensoft.net/article/46379/
https://nucet.pensoft.net/article/46379/download/pdf/
https://nucet.pensoft.net/article/46379/download/xml/
This paper does not contain new computational and experimental scientific results. It attempts to analyze, based on a simplified phenomenological approach, the methodology of the evolution histories of nuclear science and technology, as well as the contradictions and issues which, if not resolved, make senseless any discussions of scenarios for the full-scale evolution of nuclear power.
The paper analyzes in brief the evolution history of nuclear technologies in the USA and in the USSR. It also considers the present-day state of nuclear power. Two international projects, INPRO and GIF IV, were initiated in 2000. The INPRO objective is to define the evolution strategy for and the requirements to the nuclear power of tomorrow. The GIF IV project aiming to develop Generation IV reactors for future NPPs focuses on building innovative reactors capable to cope with the challenges involved in further evolution of nuclear power.
The following issues were considered as the result of the system analysis
– further evolution of nuclear power worldwide;
– nuclear non-proliferation;
– NPP safety;
– nuclear waste;
– climate and oxygen burning in the NPP operation;
– education and training of younger generations of nuclear workers.
A critical analysis into the history, status and future evolution of nuclear technologies at the present-day stage shows that the nuclear energy market has monopolized the design, development and construction of practically only one type of nuclear reactors for NPPs (95% of the NPPs under construction have water-cooled water-moderated reactors) which explains the fact that single-skilled personnel are largely trained for the construction and operation of this plant type.
Achieving the full-scale evolution level of nuclear power capable to cope with the socio-economic and ecological issues faced by humankind requires a basically new evolution concept for all fields of nuclear industry.
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en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Nuclear power
nuclear technologies
system analysis of nuclear power
Critical notes: history, state, problems and prospects of nuclear science and technology
Review Article
10.3897/nucet.5.46380
2019-09-25
nucet
Novovoronezh Nuclear Power Plant, Novovoronezh, Russia
author
Povarov, Vladimir
Novovoronezh Nuclear Power Plant, Novovoronezh, Russia
author
Fedorov, Anatoly
Novovoronezh Nuclear Power Plant, Novovoronezh, Russia
author
Vitkovsky, Sergey
2019-09-25
2019-09-25
2019
Nuclear Energy and Technology
2452-3038
5
249-256
2019
NN
Alekseenko
author
1981
1981
AD
Amaev
author
1997
1997
The concept of extending the life of power units with a VVER-440 Novovoronezh NPP.
VG
Asmolov
author
2014
text
Teploenergetika,
2014
2
16
25
2014
2014
Collection of reports (2014) Half a century of ensuring the safety of nuclear power plants with VVER in Russia and abroad. Novovoronezh NPP Publ., 521 pp. [in Russian]
BA
Gurovich
author
1997
1997
GP
Karzov
author
1993
1993
AM
Kutepov
author
1986
1986
TKh
Margulova
author
1987
1987
2017
2017
MT 1.1.4.02.1204-2017 (2017) Calculation of the resistance to brittle fracture of the WWER-440 (V-179, B-230) reactor shells, taking into account their annealing when extending the service life to 60 years. The technique. Moscow. Rosenergoatom JSC Publ., 38 pp. [in Russian]
2017
2017
MT 1.1.4.02.1205-2017 (2017) Calculation of buildings of steam generators of VVER-440 reactor installations (B-179, B-230, B-213) for resistance to brittle fracture when extending the service life to 60 years. The technique. Moscow. Rosenergoatom JSC Publ., 41 pp. [in Russian]
2016
2016
NP-001-15 (2016) General provisions for the safety of nuclear power plants. Moscow. FBU “NTTs YaRB” Publ., 57 pp. [in Russian]
2018
2018
NP-017-18 (2018) Basic requirements for extending the life of a nuclear power unit. Moscow. FBU “NTTs YaRB” Publ., 21 pp. [in Russian]
2016
2016
NP-026-16 (2016) Requirements for control systems important for the safety of nuclear power plants. Moscow. FBU “NTTs YaRB” Publ., 30 pp. [in Russian]
2015
2015
NP-096-15 (2015) Requirements for managing the life of equipment and pipelines of nuclear power plants. The main provisions. Moscow. FBU «NTTs YaRB» Publ., 19 pp. [in Russian]
FYa
Ovchinnikov
author
1988
1988
1989
1989
PNAE G-7-002-86 (1989) Standards for calculating the strength of equipment and pipelines of nuclear power plants. Moscow. Energoatomizdat Publ., 525 pp. [in Russian]
2012
2012
RD EO 0421-02 (2012) Methods for predicting the strength characteristics of the material of the reactor vessel during irradiation. Moscow. Rosenergoatom Publ., 7 pp. [in Russian]
Justification of the strength and service life of reactor vessels.
YaI
Shtrombakh
author
2006
text
Rosenergoatom (the monthly magazine for atomic energy of Russia),
2006
7
58
59
DYu
Yerak
author
2013
2013
10.3897/nucet.5.46380
https://nucet.pensoft.net/article/46380/
https://nucet.pensoft.net/article/46380/download/pdf/
https://nucet.pensoft.net/article/46380/download/xml/
The re-modernization of Unit 4 at the Novovoronezh NPP (Novovoronezh-4) made it possible to take a new approach to the problem of extending the VVER-440 reactor plant life and operation. The authors analyze the existing problems of the VVER-440/179 power unit, showing possible solutions to the identified shortcomings and the final state of the updated power unit. Modernization works significantly expanded the range of design-basis accidents from the primary coolant leak from an opening (DN = 100 mm) to the maximum possible, associated with a rupture of the main circulation pipelines (MCP) (DN = 500 mm). A unique experience was gained in using the safety systems of Unit 3, which was finally shutdown for decommissioning, to increase reliability and provide additional redundancy for the safety systems of Unit 4.
The results of the performed works showed the correctness of the adopted concept of re-extending the service life of Unit 4 and ensured its compliance with the modern safety requirements in nuclear power engineering, including as it relates to the safety impact of the first-level probabilistic safety analysis model (PSA-1) for internal initiating events.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Life re-extension
VVER-440
V-179
fluence
radiation embrittlement
safety systems
reactor vessel annealing
Some aspects of the VVER-440 reactor plant life re-extension: a case study of the Novovoronezh NPP Unit 4
Research Article
10.3897/nucet.5.46381
2019-09-25
nucet
JSC “SSC RF – IPPE n.a. A.I. Leypunsky”, Obninsk, Russia
author
Dyachenko, Petr
JSC “SSC RF – IPPE n.a. A.I. Leypunsky”, Obninsk, Russia
author
Suvorov, Alexey
JSC “SSC RF – IPPE n.a. A.I. Leypunsky”, Obninsk, Russia
author
Kukharchuk, Oleg
JSC “SSC RF – IPPE n.a. A.I. Leypunsky”, Obninsk, Russia
author
Zrodnikov, Anatoly
2019-09-25
2019-09-25
2019
Nuclear Energy and Technology
2452-3038
5
257-263
2019
10.3897/nucet.5.46381
https://nucet.pensoft.net/article/46381/
https://nucet.pensoft.net/article/46381/download/pdf/
https://nucet.pensoft.net/article/46381/download/xml/
The concept of a high power reactor-laser system based on a nuclear pumped optical quantum amplifier (OKUYaN) was formulated at IPPE in the mid-1980-ies. The idea amounted to the use of wide-aperture OKUYaN as an amplifier within the already well-known “master laser – two-pass amplifier with phase conjugation” scheme.
The structure of such an amplifier includes a system of two neutron-coupled units – an ignition reactor (RB) and a nuclear pumped laser amplifier (LB). The ignition unit is a compact multi-core pulsed fast neutron reactor. The laser amplifier unit operates on thermal neutrons and, with regard to the neutronics, it is a subcritical booster zone of the ignition reactor unit.
Unique reactor-laser complex incorporating demonstration sample of a pulsed reactor-laser system based on OKUYaN (test facility “Stand B”) having no analogues anywhere in the world, was developed and put into operation at IPPE in 1999 for the purpose of substantiation of basic principles of the OKUYaN concept and demonstration of the possibility of its practical implementation, as well as verification of calculation codes and development of relevant equipment elements.
Problems overcome in the development and construction of “Stand B” test facility, the choice and justification of the neutronics and laser characteristics of the OKUYaN demonstration sample are discussed in the present paper. Provided are the results of a detailed computational-experimental study of the demonstration sample characteristics, the data from systems studies of direct conversion of nuclear fission energy into laser radiation energy in complex reactor-laser devices and the results of examination of prospects for the development of nuclear-laser power engineering.
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en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Multi-core nuclear reactor
neutrons
fission fragments
nuclear pumping
laser
optical quantum amplifier with nuclear pumping
Problem of nuclear-laser power engineering and methods of their solution
Research Article
10.3897/nucet.5.46382
2019-09-25
nucet
Rosenergoatom Concern, Moscow, Russia
author
Petrov, Andrey
Rosenergoatom Concern, Moscow, Russia
author
Shutikov, Alexander
Rosenergoatom Concern, Moscow, Russia
author
Ponomarev-Stepnoy, Nikolay
Rosenergoatom Concern, Moscow, Russia
author
Bezzubtsev, Valery
Rosenergoatom Concern, Moscow, Russia
author
Bakanov, Mikhail
Rosenergoatom Concern, Moscow, Russia
author
Troyanov, Vladimir
2019-09-25
2019-09-25
2019
Nuclear Energy and Technology
2452-3038
5
265-271
2019
PN
Alekseev
author
2016
Two-component nuclear power system with thermal and fast reactors in the closed nuclear fuel cycle. Ed. by the academician of RAS Ponomarev-Stepnoy NN.
2016
160 pp
AK
Bhaduri
author
2017
2017
2017
2017
Executive order of the Government of the Russian Federation No. 1209-r (2017) on the approval of the General layout of arrangement of power generation facilities until 2035. http://static.government.ru/media/files/zzvuuhfq2f3OJIK8AzKVsXrGIbW8ENGp.pdf [accessed May 12, 2019, in Russian]
F
Heidet
author
2019
2019
DA
Klinov
author
2018
Two-component nuclear power with the closed fuel cycle and a role of reactors on thermal and fast neutrons
2018
F
Laugier
author
2019
2019
AYu
Petrov
author
2018
A role and the place of operating organization in implementation of the strategy of development of nuclear power industry of Russia
2018
S
Pivet
author
2017
2017
Y
Sagayama
author
2017
2017
AV
Shutikov
author
2017
2017
Hui
Zhang Dong
author
2017
2017
10.3897/nucet.5.46382
https://nucet.pensoft.net/article/46382/
https://nucet.pensoft.net/article/46382/download/pdf/
https://nucet.pensoft.net/article/46382/download/xml/
Possible options of organization of two-component energy system with closed nuclear fuel cycle (CNFC) and new business potential associated with provision of CFC services to foreign customers are examined.
Dominating role in the development of nuclear power generation is assigned to VVER reactors with gradually increasing fraction of sodium-cooled fast breeder reactors (FR) incorporated within the joint nuclear fuel cycle operated on MOX-fuel.
Components of such energy system perform the following functions:
1. Fast reactors:
Generate electric power in base-load mode (possibility of fine tuning of reactor power within limited range (100 – 75 – 100%) is assumed);
Utilize waste and/or regenerated uranium for re-fueling power reactors, produce plutonium applicable to the maximum extent for manufacturing MOX-fuel for VVER reactors;
Incinerate long-lived highly radioactive wastes – minor actinides separated during reprocessing spent nuclear fuel of FR and VVER reactors.
2. VVER reactors:
Generate electricity in compliance with manoeuvrability requirements imposed by the utility company operating the energy system;
Utilize MOX-fuel instead of UO2 fuel;
Are offered for export with the option of returning SNF back to Russia;
Plutonium extracted from VVER spent fuel is used for manufacturing MOX-fuel for SFR.
3. Nuclear fuel cycle facilities:
Provide reprocessing of SNF from VVER and SFR with extraction of nuclear materials for recycling;
Use depleted or reprocessed uranium and plutonium extracted from spent nuclear fuel for manufacturing MOX-fuel;
Provide partitioning of RAW for subsequent utilization of minor actinides and reduction of risks of proliferation of nuclear materials, conditioning and disposal of RAW.
Russia possesses capacities for establishing the two-component system with CNFC, as well as the new business approach to rendering CNFC services to foreign customers.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Two-component nuclear energy system
centralized nuclear fuel cycle
commercial power supply complex of the Supplier (Rosatom)
Prospects of creation of the two-component nuclear energy system
Research Article
10.3897/nucet.5.39226
2019-09-25
nucet
Institute for Nuclear Power Engineering, Obninsk, Russia
author
Telnov, Victor
https://orcid.org/0000-0003-0176-5016
Institute for Nuclear Power Engineering, Obninsk, Russia
author
Korovin, Yury
2019-09-25
2019-09-25
2019
Nuclear Energy and Technology
2452-3038
5
273-280
2019
10.3897/nucet.5.39226
https://nucet.pensoft.net/article/39226/
https://nucet.pensoft.net/article/39226/download/pdf/
https://nucet.pensoft.net/article/39226/download/xml/
The technologies of knowledge representation and inference in an artificial intelligence system focused on the domain of nuclear physics and nuclear power engineering are considered. The possibilities of description logics and graph databases of nuclear knowledge for the generation of cognitive hypotheses, using in addition to deduction and other ways of reasoning, such as inductive inference and reasoning based on analogies, are discussed. The use of adequate description logic and measures of semantic similarity is substantiated. Interactive visual navigation and reasoning on the knowledge graphs are performed by means of special retrieval widgets and the smart RDF browser. Operations with semantic repositories are implemented on cloud platforms using SPARQL queries and RESTful services. The proposed software solutions are based on cloud computing using DBaaS and PaaS service models to ensure scalability of data warehouses and network services. Example of use of the offered technologies and software has been given.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Nuclear education
semantic web
knowledge graph
cloud computing
Semantic web and knowledge graphs as an educational technology of personnel training for nuclear power engineering
Research Article
10.3897/nucet.5.47972
2019-12-10
nucet
Ural Federal University named after the First President of Russia B. N. Yeltsin, Yekaterinburg, Russia
University of Technology, Baghdad, Iraq
author
Abed, Akram Hamzah
Ural Federal University named after the First President of Russia B. N. Yeltsin, Yekaterinburg, Russia
author
Shcheklein, Sergey
Ural Federal University named after the First President of Russia B. N. Yeltsin, Yekaterinburg, Russia
author
Pakhaluev, Valery
2019-12-10
2019-12-10
2019
Nuclear Energy and Technology
2452-3038
5
281-287
2019
10.3897/nucet.5.47972
https://nucet.pensoft.net/article/47972/
https://nucet.pensoft.net/article/47972/download/pdf/
https://nucet.pensoft.net/article/47972/download/xml/
Advanced nuclear power plants are equipped with passive emergency heat removal systems (PEHRS) for removing the decay heat from reactor equipment in accidents accompanied by primary circuit leakage to the final heat absorber (ambient air). Herein, the intensity of heat dissipation to air from the outer surface of the heat exchanger achieved by buoyancy induced natural convection is extremely low, which need to a large heat exchanger surface area, apply different types of heat transfer intensification including (grooves, ribs and extended surfaces, positioning at higher altitudes, etc.). The intensity of heat removal is also strongly dependent on the ambient air temperature (disposable temperature head).
Construction of nuclear power plants in countries with high ambient temperatures (Iran, Bangladesh, Egypt, Saudi Arabia, and others) which are characterized by a high level of ambient temperature imposes additional requirements on the increase of the heat exchange surfaces.
The experimental investigation results of heat transfer intensification by a low energy ultrasonic which supply a fine liquid droplet (size ~3 µm) in the cooling air are presented in the present paper. In such case, the heat transfer between the surface and cooling flow involves the following three physical effects: convection, conductive heat transfer, and evaporation of water droplets. The last two effects weakly depend on the ambient air temperature and provide an active heat removal in any situation.
The investigation was performed using a high-precision calorimeter with a controlled rate of heat supply (between 7800 and 12831 W/m2) imitating heated surface within the range of Reynolds numbers from 2500 to 55000 and liquid (water) flow rates from 23.39 to 111.68 kg·m-2·h-1.
The studies demonstrated that the presence of finely dispersed water results in a significant increase in heat transfer compared with the case of using purely air-cooling. With a fixed heat flux, the energy efficiency increases with increasing water concentration, reaching the values over 600 W·m-2·C-1 at 111.68 kg·m-2·h-1, which is 2.8 times higher than for air cooling. With further development of research in order to clarify the optimal areas of intensification, it is possible to use this technology to intensify heat transfer to the air in dry cooling towers of nuclear power plants and thermal power plants used in hot and extreme continental climates.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Nuclear power plants
aerosol cooling
air-water mist
water concentration
heat exchange intensification.
Heat transfer intensification in emergency cooling heat exchanger and dry cooling towers on nuclear power plant using air-water mist flow
Research Article
10.3897/nucet.5.48391
2019-12-10
nucet
Engineering School of Nuclear Technology «National Research Tomsk Polytechnic University», Tomsk, Russia
author
Bulakh, Olga
Engineering School of Nuclear Technology «National Research Tomsk Polytechnic University», Tomsk, Russia
author
Kostylev, Oleg
Engineering School of Nuclear Technology «National Research Tomsk Polytechnic University», Tomsk, Russia
author
Nesterov, Vladimir
Engineering School of Nuclear Technology «National Research Tomsk Polytechnic University», Tomsk, Russia
author
Cherdizov, Eldar
2019-12-10
2019-12-10
2019
Nuclear Energy and Technology
2452-3038
5
289-295
2019
10.3897/nucet.5.48391
https://nucet.pensoft.net/article/48391/
https://nucet.pensoft.net/article/48391/download/pdf/
https://nucet.pensoft.net/article/48391/download/xml/
High-temperature gas-cooled reactor (HTGR) is one of promising candidates for new generation of nuclear power reactors. This type of nuclear reactor is characterized with the following principal features: highly efficient generation of electricity (thermal efficiency of about 50%); the use of high-temperature heat in different production processes; reactor core self-protection properties; practical exclusion of reactor core meltdown in case of accidents; the possibility of implementation of various nuclear fuel cycle options; reduced radiation and thermal effects on the environment, forecasted acceptability of financial performance with respect to cost of electricity as compared with alternative energy sources.
The range of output coolant temperatures in high-temperature reactors within the limits of 750–950 °C predetermines the use of graphite as the structural material of the reactor core and helium as the inert coolant. Application of graphite ensures higher heat capacity of the reactor core and its practical non-meltability.
Residence time of reactor graphite depends on the critical value of fluence of damaging neutrons (neutrons with energies above 180 keV). In its turn, the value of critical neutron fluence is determined by the irradiation temperature and flux density of accompanying gamma-radiation. The values of critical fluence for graphite decrease within high-temperature region of 800–1000 °C to 1·1022 – 2·1021 cm–2, respectively. The compactness of the core results in the increase of the fracture of damaging neutrons in the total flux. These circumstances predetermine relatively low values of lifespan of graphite structures in high-temperature reactors.
Design features and operational parameters of GT-MHR high-temperature gas-cooled reactor are described in the present paper. Results of neutronics calculations allowing determining the values of damaging neutron flux, nuclear fuel burnup and expired lifespan of graphite of fuel blocks were obtained. The mismatch between positions of the maxima in the dependences of fuel burnup and exhausted lifespan of graphite in fuel blocks along the core height is demonstrated.
The map and methodology for re-shuffling fuel blocks of the GT-MHR reactor core were developed as the result of analysis of the calculated data for ensuring the matching between the design value of the fuel burnup and expected total graphite lifespan.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Critical fluence
reactor-grade graphite
damaging neutrons
graphite lifespan
nuclear fuel burnup
GT-MHR
HTGR
Extension of lifespan of graphite in fuel blocks of high-temperature gas-cooled reactors as the resource for ensuring design values of nuclear fuel burn-up
Research Article
10.3897/nucet.5.48392
2019-12-10
nucet
Branch of MPEI in Volzhsky, Volzhsky, Russia
author
Kuzevanov, Vyacheslav
Branch of IDGC of the South, PJSC, Volzhsky, Russia
author
Podgorny, Sergey
2019-12-10
2019-12-10
2019
Nuclear Energy and Technology
2452-3038
5
297-303
2019
J
Anderson
author
2009
2009
Canonsburg.
2016
text
ANSYS, inc.
2016
247
.
Canonsburg.
2016a
text
ANSYS, inc.
2016a
100
.
Canonsburg.
2016b
text
ANSYS, inc.
2016b
177
.
2005
2005
Design of the Reactor Core for Nuclear Power Plants (2005) Safety Guide No. NS-G-1.12. Vienna. International Atomic Energy Agency: 3–8.
10.2172/5176415
10.2172/7283150
2002
2002
GT-MHR Conceptual Design Description Report (2002) NRC project No. 716. San Diego. General Atomics: 58–62.
2014
2014
International Safeguards in the Design of Nuclear Reactors (2014) IAEA Nuclear Energy Series No. NP-T-2.9. Vienna. International Atomic Energy Agency: 18–23.
10.1088/1742-6596/891/1/012069
10.26583/npe.2018.4.03
10.17588/2072-2672.2017.4.021-029
B
Mohammadi
author
1994
1994
AJ
Neylan
author
1994
1994
T
Petrila
author
2005
2005
SK
Podgorny
author
2017
2017
2014
2014
Safety of Nuclear Power Plants: Design (2014) Specific safety requirements No. SSR-2/1 (Rev.1). Vienna. International Atomic Energy Agency: 4–10.
CT
Shaw
author
1992
1992
A
Vasyaev
author
2001
2001
10.3897/nucet.5.48392
https://nucet.pensoft.net/article/48392/
https://nucet.pensoft.net/article/48392/download/pdf/
https://nucet.pensoft.net/article/48392/download/xml/
Positive effect of profiling the gas-cooled reactor core within the framework of the GT-MHR project was investigated in (Podgorny and Kuzevanov 2017, Kuzevanov and Podgorny 2017, 2018). The necessity arises to supplement already implemented analysis of equilibrium conditions of core operation with investigation of effects of profiling on the temperature field in transient modes of reactor core operation.
The present paper is dedicated to the investigation of development of transients in gas-cooled nuclear reactor core subject to the implementation of different principles of core profiling.
Investigation of transients in reactor core represents complex problem, solution of which by conducting direct measurements is beyond the resources available to the authors. Besides the above, numerical simulation based on advanced CFD software complexes (ANSYS 2016, 2016a, 2016b, Shaw 1992, Anderson et al. 2009, Petrila and Trif 2005, Mohammadi and Pironneau 1994) is also fairly demanding in terms of required computer resources.
The algorithm for calculating temperature fields using the model where the reactor core is represented as the solid medium with gas voids was developed by the authors and the assumption was made that heat transfer due to molecular heat conductivity can be described by thermal conductivity equation written for continuous medium with thermal physics parameters equivalent to respective parameters of porous object in order to get the possibility of obtaining prompt solutions of this type of problems.
Computer code for calculating temperature field in gas-cooled reactor in transient operation modes was developed based on the suggested algorithm. Proprietary computation code was verified by comparing the results of numerous calculations with results of CFD-modeling of respective transients in the object imitating the core of gas-cooled nuclear reactor. The advantage of the developed computer code is the possibility of real-time calculation of evolution of conditions in complex configurations of gas-cooled reactor cores with different channel diameters. This allows using the computer code in the calculations of transients in loops of reactor facility as a whole, in particular for developing reactor simulators.
Results are provided of calculations of transients for reactor core imitating the core of gas-cooled nuclear reactor within the framework of GT-MHR project performed for different approaches to profiling coolant mass flow. Results of calculations unambiguously indicate the significant difference of temperature regimes during transients in the reactor core with and without profiling and undeniable enhancement of reliability of nuclear reactor (Design of the Reactor Core 2005, International Safeguards 2014, Safety of Nuclear Power Plants 2014) with profiling of coolant mass flow in the reactor core as a whole.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Reactor core profiling
transient processes
temperature field
porous body
thermal conductivity equation
heat sink
Temperature field in gas-cooled reactor core in transient conditions under different approaches to mass flow profiling
Research Article
10.3897/nucet.5.48393
2019-12-10
nucet
Obninsk Institute for Nuclear Power Engineering, Obninsk, Russia
author
Bragin, Iliya
Obninsk Institute for Nuclear Power Engineering, Obninsk, Russia
author
Belozerov, Vladimir
2019-12-10
2019-12-10
2019
Nuclear Energy and Technology
2452-3038
5
305-311
2019
10.3897/nucet.5.48393
https://nucet.pensoft.net/article/48393/
https://nucet.pensoft.net/article/48393/download/pdf/
https://nucet.pensoft.net/article/48393/download/xml/
To simulate the mode of the RCP starting in an earlier inoperative loop, KORSAR/GP, a code supporting coupled numerical modeling of neutronic and thermal-hydraulic transients in a VVER reactor plant in operating and emergency conditions, was chosen as the computational tool.
Studying these modes using thermal-hydraulic codes makes it possible to analyze the course of transients and certain emergency processes without using commercial test procedures, which contributes to laying the groundwork for addressing the issues involved in ensuring the reliability, operating safety and efficiency of nuclear power plants.
Increased requirements to the safety of NPPs identify the need for avoiding excessive conservatism in the analysis based on which requirements to safety systems are formulated, as well as for enhancing the knowledge of the regularities of thermal-hydraulic transients based on advanced computer programs (or codes) designed for improved computational analysis of non-stationary thermal hydraulics in the water-cooled reactor circulation circuits in emergency and transient modes relying on inhomogeneous non-equilibrium mathematical models of two-phase flows and on a detailed description of the physical transient regularities.
The purpose of the study is to analyze computationally the starting of a VVER-1000 RCP in an earlier inoperative loop at different reactor plant power values. To do this, one requires to develop the VVER-1000 reactor primary circuit computational pattern to model the transient taking place as one RCP is started, to conduct a further analysis, and to compare the key monitored reactor coolant and core parameters (power, temperature, flow rate, etc.).
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Mode
reactor coolant pump (RCP)
circulation loop
reactor plant
scram
reactivity coefficient
safety margin
reactor plant power
A study into the modes of the VVER-1000 RCP starting in an earlier inoperative loop
Research Article
10.3897/nucet.5.48394
2019-12-10
nucet
Obninsk Institute for Nuclear Power Engineering, Obninsk, Russia
author
Trofimov, Maksim
Obninsk Institute for Nuclear Power Engineering, Obninsk, Russia
author
Globa, Ruslan
2019-12-10
2019-12-10
2019
Nuclear Energy and Technology
2452-3038
5
313-316
2019
10.3897/nucet.5.48394
https://nucet.pensoft.net/article/48394/
https://nucet.pensoft.net/article/48394/download/pdf/
https://nucet.pensoft.net/article/48394/download/xml/
During operation of nuclear reactors, there are various factors that affect the nuclear plant piping leading to erosion of the pipe inner surface and an increase of its micro-relief (roughness). Metal corrosion occurs and spreads faster on a surface having a higher value of the roughness parameter. Failure through erosive wear of the parent metal takes place predominantly in the pipe bend area. The roughness of the pipe inner surface has a sizeable effect on the signal attenuation in the process of the pipe wall ultrasonic testing. Defective main pipeline segments were cut out during preventive repairs from which samples with different operating times were taken. Five defective pipe segments of the austenitic 12Kh18N10T grade steel cut out of a high-pressure reheater’s piping and five defective pipe segments of the perlite-class steel of grade 20, after different operating times, were used to determine experimentally the actual value of the pipe inner surface roughness. Besides, a piece of a new Ø273 × 12 pipe of the 12Kh18N10T steel and a piece of a Ø159 × 6 pipe of grade 20 steel were cut out. The inner surface roughness was measured for different segments. Dependences of the roughness value on the operating time and the pipe segment type have been obtained. Company specimens were fabricated with the inner surfaces having the roughness corresponding to various pipe operating times. This made it possible to take into account the influence of the inner surface roughness on the signal attenuation in the process of the weld integrity ultrasonic testing and during ultrasonic measurements of the weld adjacent zone grain size value following the weld repair.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Life extension
surface roughness
metal corrosion
operating conditions
Elcometer 7061 Marsurf PS1 roughness meter
parent metal
preventive maintenance
Roughness of the nuclear reactor pipe inner surface depending on the reactor operating time
Research Article
10.3897/nucet.5.48397
2019-12-10
nucet
National Research Nuclear University, Obninsk, Russia
author
Doan, Thi Zieu Chang
National Research Nuclear University, Obninsk, Russia
author
Lazarenko, Georgy
National Research Nuclear University, Moscow, Russia
author
Lazarenko, Denis
2019-12-10
2019-12-10
2019
Nuclear Energy and Technology
2452-3038
5
317-321
2019
Use of the VVR-SM Reactor for the Development of the Neutron Capture Therapy Method in Uzbekistan. Proceedings of the Russian Academy of Sciences.
GA
Abdullaeva
author
2009
text
Series: Physics
2009
73
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540
544
АА
Andrianov
author
2012
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2012
180 pp
GA
Bat’
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1985
Research Nuclear Reactors.
1985
280 pp
10.15688/jvolsu4.2016.1.11
VМ
Budov
author
1985
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1985
264 pp
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Chirkin
author
1968
Thermophysical Properties of Nuclear Technique Materials. Handbook.
1968
484 pp
BA
Dementiev
author
1990
Nuclear Power Reactors.
1990
352 pp
PL
Kirillov
author
2010
Handbook for Thermohydraulic Calculations in Nuclear Power.
2010
770 pp
PL
Kirillov
author
2000
Heat and Mass Transfer in Nuclear Power Plants.
2000
456 pp
PL
Kirillov
author
1990
Handbook for Thermohydraulic Calculations (Nuclear Reactors, Heat Exchangers, Steam Generators).
1990
360 pp
RP
Kuatbekov
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Kutateladze
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1990
Heat Transfer and Hydraulical Resistance. Reference book.
1990
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Novozhilova
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of NNSTU
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Petukhov
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1967
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1967
411 pp
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Petukhov
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Heat Transfer in Nuclear Power Plants.
2003
548 pp
1991
Ed. Grigoryeva IS, Meilikhova EZ.
1991
1232 pp
Physical Quantities: Reference (1991) Ed. Grigoryeva IS, Meilikhova EZ.Energoatomizdat Publ., Мoscow, 1232 pp. [in Russian]
The Optimal Distribution of the Irradiation Resource of a Research Reactor in Large-Scale Production of Radionuclides. Col.
VA
Tarasov
author
2018
text
papers of JSC SSC RIAR
2018
1
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Tretyakov
author
2014
2014
Development of Projects for Perspective Basin Research Reactors. VANT. Ser.
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Tretyakov
author
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Obespechenie Bezopasnosti AES [NPP safety]
2013
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Udaev
author
1973
Heat Transfer.
1973
360 pp
10.3897/nucet.5.48397
https://nucet.pensoft.net/article/48397/
https://nucet.pensoft.net/article/48397/download/pdf/
https://nucet.pensoft.net/article/48397/download/xml/
Having thoroughly analyzed the design features of VVER-type pressurized water reactors and VVR-type research reactors, the authors propose a design of a research reactor with low-enriched fuel based on deeply updated VVER-440 fuel assemblies. The research reactor is intended to solve a wide range of applied problems in nuclear physics, radiation chemistry, materials science, biology, and medicine. The calculated thermal hydraulics confirms the correctness of the fundamental approaches laid down in the reactor design.
An equivalent reactor core model in the form of a thick-walled cylinder was considered, and the radial power density distribution was obtained. According to the heat power level, five groups of FAs were identified. For each group, the coolant mass flow rate was calculated, which ensures alignment with the outlet coolant temperature.
The coolant flow regime was also estimated. It turned out that for the first row of FAs, the flow regime is in the transition region, while for the other rows the flow regime is laminar. A test by the Gr.Pr≥1.105 criterion showed its conformity (the calculated value was 1.96.106), indicating the transition to a viscous-gravitational regime. The FE surface overheating was calculated relative to the mixed coolant average temperature. The axial coolant flow temperature distribution is the same in all the FAs, the change in power is compensated by the corresponding change in the coolant flow. The maximum coolant overheating on the FE wall relative to the flow core is observed in the central FAs, reaching 31 °C, the boiling margin is about 15 °C.
The estimates showed a significant dynamic pressure margin during natural thermal-convective circulation. By calculation, the values of the FE surface overheating during the reactor normal operation were obtained. An approximately 15-degree surface overheating margin relative to the saturation curve is shown, which guarantees the absence of cavitation wear of the FE claddings. In general, the performed calculations confirmed the correctness of the approaches laid down in the reactor design and made it possible to specify the core thermal hydraulics necessary for further developing the concept.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Research reactor
low-enriched fuel
natural circulation
long-term campaign
export potential
VVR
IRT
VVER-440
Calculations of research reactor thermal hydraulics based on VVER-440 fuel assamblies
Research Article
10.3897/nucet.5.48423
2019-12-10
nucet
Research Institute of Nuclear Materials, Zarechny, Russia
author
Barsanova, Svetlana
Research Institute of Nuclear Materials, Zarechny, Russia
author
Kinev, Evgeniy
Research Institute of Nuclear Materials, Zarechny, Russia
author
Kozlov, Aleksander
Research Institute of Nuclear Materials, Zarechny, Russia
author
Portnykh, Irina
Research Institute of Nuclear Materials, Zarechny, Russia
author
Panchenko, Valeriy
Research Institute of Nuclear Materials, Zarechny, Russia
author
Evseev, Mikhail
2019-12-10
2019-12-10
2019
Nuclear Energy and Technology
2452-3038
5
323-329
2019
10.3897/nucet.5.48423
https://nucet.pensoft.net/article/48423/
https://nucet.pensoft.net/article/48423/download/pdf/
https://nucet.pensoft.net/article/48423/download/xml/
The swelling, corrosion and high-temperature embrittlement behavior of the fast-neutron sodium-cooled reactor standard and test fuel rod claddings was studied following the operation up to a damaging dose of 55 to 69 dpa. The tested characteristics were found to differ sensitively in conditions similar to irradiation for the claddings of the experimental tube conversion technology. Unlike the standard fuel rod claddings, the test rod claddings were additionally heated in the process of fabrication to homogenize the solid solution at different temperatures and austenitization times. On the whole, this led to an increased cladding resistance contrary the damaging factor of the reactor environment. The positive effect is explained by the influence of carbon and the morphology of swelling-reducing alloying elements, as well as by the nature of the carbide and intermetallide phase precipitation. However, the dispersion of the post-irradiation properties which remained significant and was also earlier observed in the standard rods is explained by potential differences in the heat treatment technology and the irradiation temperature in conditions of a hard-to-control coolant flow velocity. The swelling rate and the in-fuel corrosion depth for the test technology tubes were respectively 0.04 to 0.058%/dpa and 20 to 47 μm; similar values for the test material are 0.036 to 0.056%/dpa and 15 to 35 μm respectively. The short-term mechanical properties of the test fuel rods at a temperature of 600 °C showed a smaller tendency towards high-temperature embrittlement. The dispersion of the properties was caused by the chemical and structural heterogeneity as the result of the tube fabrication.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Fuel rod
cladding
swelling rate
corrosion
grain size
high-temperature embrittlement
manufacturing technology.
Analyzing the causes for the dispersion of the fast reactor spent fuel rod cladding properties
Research Article
10.3897/nucet.5.48425
2019-12-10
nucet
JSC “SSC RF – IPPE” n.a. A.I. Leypunsky, Obninsk, Russia
author
Ivanov, Igor
JSC “SSC RF – IPPE” n.a. A.I. Leypunsky, Obninsk, Russia
author
Shelemetyev, Vasily
JSC “SSC RF – IPPE” n.a. A.I. Leypunsky, Obninsk, Russia
author
Askhadullin, Radomir
2019-12-10
2019-12-10
2019
Nuclear Energy and Technology
2452-3038
5
331-336
2019
10.3897/nucet.5.48425
https://nucet.pensoft.net/article/48425/
https://nucet.pensoft.net/article/48425/download/pdf/
https://nucet.pensoft.net/article/48425/download/xml/
As part of the project on developing methods for removing hydrogen and tritium from the circulation circuits of reactor plants with heavy liquid metal coolants, the authors studied the kinetics of bismuth oxide reduction by hydrogen in the temperature range of 425–500 °C and hydrogen concentrations of 25–100 vol.%. The kinetic characteristics of the test reaction were determined by continuous measurements of the water steam (reaction product) concentration in mixtures of hydrogen with helium that passed through a heated reaction vessel with a sample of bismuth oxide. The water steam concentration was measured by a thermal-conductivity detector. The obtained time dependences of the bismuth oxide reduction degree (with varying reaction conditions) were processed by the affine time transformation method. It was also found that the reduction process ran in kinetic mode. The reduction mechanism is the same in the entire temperature range. The limiting reaction stage is the adsorption of hydrogen on the surface of the bismuth oxide sample. The time dependence of the reduction degree is in good agreement with Avrami-Erofeev equation with n = 1. The reaction activation energy is 92.8 ± 1.9 kJ/mol. The reduction reaction rate is directly proportional to the concentration of hydrogen in its mixture with an inert gas.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Heavy liquid metal coolants (HLMC)
nuclear safety
hydrogen
tritium
hydrogen purification of HLMC
hydrogen afterburner
bismuth oxide
reaction kinetics
thermal-conductivity analysis
affine time transformation method
Avrami-Erofeev equation.
A study on the kinetics of bismuth oxide reduction by hydrogen as applied to the technology of removing hydrogen from circulation circuits with heavy liquid metal coolants
Research Article
10.3897/nucet.5.48427
2019-12-10
nucet
JSC “SSC RF – IPPE” n.a. A.I. Leypunsky, Obninsk, Russia
author
Ivanov, Sergey
JSC “SSC RF – IPPE” n.a. A.I. Leypunsky, Obninsk, Russia
author
Porollo, Sergey
JSC “SSC RF – IPPE” n.a. A.I. Leypunsky, Obninsk, Russia
author
Baranaev, Yury
JSC “SSC RF – IPPE” n.a. A.I. Leypunsky, Obninsk, Russia
author
Timofeev, Vladimir
JSC “SSC RF – IPPE” n.a. A.I. Leypunsky, Obninsk, Russia
author
Kharizomenov, Yury
2019-12-10
2019-12-10
2019
Nuclear Energy and Technology
2452-3038
5
337-343
2019
10.3897/nucet.5.48427
https://nucet.pensoft.net/article/48427/
https://nucet.pensoft.net/article/48427/download/pdf/
https://nucet.pensoft.net/article/48427/download/xml/
Spent nuclear fuel (SNF) storage in reactor spent fuel pools (SFP) is one of the crucial stages of SNF management technology: it requires special measures to ensure nuclear and radiation safety. During long-term storage in water-filled SFPs, leak-tight canisters in which SFAs are usually placed can become unsealed, which will result in the development of corrosion processes in the fuel element (FE) claddings.
We studied fragments of spent fuel elements of the AM reactor of the World’s First NPP during their long exposure in the aqueous medium. The aim of the study was to obtain experimental data on the corrosion changes in the FE claddings and fuel composition during storage as well as on the release of radioactive fission products from them. For the study, a laboratory facility for exposing fuel elements in the water was developed and experimental fragments of fuel elements were made. The study was carried out in the hot chamber of the SSC RF-IPPE. The change in the activity of the water was estimated by the γ-dose rate from the selected water sample. The diameter measurements and metallographic studies were carried out in various sections of FE fragments.
Corrosion tests were carried out on fragments of spent fuel elements of the AM reactor of the World’s First NPP that were stored for a long time (more than 50 years – FEs with U-Mo fuel and ~ 20 years – FEs with UO2 fuel) using standard technology – first in SFP canisters filled with water and then in dry canisters in the air. Placing the fuel elements in the water did not lead to through damage to the FE claddings and a significant change in the size (diameter) of the outer cladding. Metallographic studies of the FE fragments after the corrosion tests showed the presence of intergranular and local frontal corrosion on the surface of the claddings, the depth of which exceeded the depth of the cladding corrosion defects before testing. The rate of radionuclide release from the fuel composition was estimated by the γ-dose rate of water samples taken from the glasses with FE fragments. Throughout the test period, the dose rate of water samples from the glasses with defect-free FEs remained constant. The dose rate from water samples taken from the glasses with the FE fragments with an artificial defect grew during storage.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Fuel pin
fuel assembly (FA)
spent fuel assembly (SFA)
corrosion
defect
aquatic environment
artificial defect
fuel composition
metallographic studies
γ-dose rate
material microstructure
fuel element (FE) claddings
Bilibino NPP
Corrosion tests in water of fuel elements irradiated in the world’s first NPP reactor
Research Article
10.3897/nucet.5.48426
2019-12-10
nucet
Obninsk Institute for Nuclear Power Engineering, Obninsk, Russia
author
Kazansky, Yury
Obninsk Institute for Nuclear Power Engineering, Obninsk, Russia
author
Karpovich, Gleb
2019-12-10
2019-12-10
2019
Nuclear Energy and Technology
2452-3038
5
345-351
2019
10.3897/nucet.5.48426
https://nucet.pensoft.net/article/48426/
https://nucet.pensoft.net/article/48426/download/pdf/
https://nucet.pensoft.net/article/48426/download/xml/
Simulating fast neutron reactor cores for comparing experimental and calculated data on the reactor neutronics characteristics is performed using zero power test stands. The BFS test facilities in operation in Russia (Obninsk) are discussed in the present paper. The geometrical arrangement of materials in the cores of the simulated reactors (fuel pins, fuel assemblies, coolant geometry) differs from the simulation assembly on the BFS. This can cause differences between the experimental results obtained at the BFS and theoretical calculations even in the case when homogenized concentrations of all materials of the reactor are thoroughly observed. The resulting differences in neutronics parameters due to the geometry of arrangement of materials with the same homogeneous concentrations are referred to as the heterogeneous effect. Heterogeneous effects tend to increase with increasing reactor power and its size, mainly due to changes in the neutron spectra.
Calculations of a number of functional values were carried out for assessing the heterogeneous effects for different spatial arrangements of the reactor materials. The calculations were performed for the following cases: a) heterogeneous distribution of materials in accordance with the design of a fast reactor; b) heterogeneous arrangement of materials in accordance with the capabilities and design features of the BFS test facility; c) homogeneous representation of materials in the reactor core and breeding blankets.
The configuration of materials in accordance with the design data for fast reactors of the BN-1200 type was accepted as the basic calculation option, relative to which the effect called the heterogeneous shift of the functional value (HSF) was calculated. The effect of neutron leakage on the HSF obtained as the result of calculations using different boundary conditions was estimated. All calculations were carried out for the same homogeneous concentrations of all materials for all the above three configurations. Calculations were carried out as well for the case when plutonium metal fuel was used in the BFS.
The values of the following functionals were calculated for different cases of arrangement of materials: the effective multiplication factor (reactivity), the sodium void reactivity effect, the average energy of fission-inducing neutrons, and the ratios of radioactive capture cross-sections to fission cross-sections for 239Pu. The calculations were performed using the Serpent 2.1.30 (VTT, Finland) Monte Carlo software package for neutronics simulations and ENDF/B-VII.0 and JEFF-3.1.1 evaluated nuclear data libraries.
The effects of various options of material arrangement on the values of keff were found to be the greatest (about 1.6%) for the case when fissile material in the form of dioxide is replaced with metal fissile material. Homogenization of the composition reduces the keff value by about 0.4%.
The average energy of fission-inducing neutrons depends to a significant extent on the leakage of neutrons and the presence of sodium (the average energy of neutrons increases and reaches in the presence of sodium about 100 keV, that is, it increases by about 11–13%). Replacing fissile material metal with its dioxide in the BFS test facility (while maintaining homogeneous concentrations, including that of oxygen) allows reducing the average energy of fission-inducing neutrons by about 60 keV.
The highest values of HSF, reaching 65%, are observed when calculation of sodium void reactivity effect is performed with materials distributed homogeneously; however, HSF is equal to 1.5% when calculation of the reactor mock-up assembled on the BFS is performed. In the absence of neutron leakage (infinitely extended medium), the sodium void reactivity effect becomes positive and the HSF is equal to 4–7%.
The heterogeneous effect of α for 239Pu noticeably (6–8%) depends only on the replacement of metallic plutonium with its dioxide (maintaining, of course, the homogeneous concentrations).
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Heterogeneous effect
simulation of nuclear reactors
critical test benches
fast reactor
BN
sodium void reactivity effect
neutronics calculations
Monte-Carlo
Heterogeneous effects in simulating a fast nuclear reactor on the BFS test facility
Research Article
10.3897/nucet.5.46517
2019-12-10
nucet
Innovative and Technology Center, Moscow, Russia
author
Egorov, Alexander
Innovative and Technology Center, Moscow, Russia
author
Khomyakov, Yurii
Innovative and Technology Center, Moscow, Russia
author
Rachkov, Valerii
Innovative and Technology Center, Moscow, Russia
author
Rodina, Elena
https://orcid.org/0000-0002-1253-7708
Innovative and Technology Center, Moscow, Russia
author
Suslov, Igor
2019-12-10
2019-12-10
2019
Nuclear Energy and Technology
2452-3038
5
353-359
2019
10.3897/nucet.5.46517
https://nucet.pensoft.net/article/46517/
https://nucet.pensoft.net/article/46517/download/pdf/
https://nucet.pensoft.net/article/46517/download/xml/
The Russian Federation is developing a number of technologies within the «Proryv» project for closing the nuclear fuel cycle utilizing mixed (U-Pu-MA) nitride fuel. Key objectives of the project include improving fast reactor nuclear safety by minimizing reactivity changes during fuel operating period and improving radiological and environmental fuel cycle safety through Pu multi-recycling and МА transmutation.
This advanced technology is expected to allow operating the reactor in an equilibrium cycle with a breeding ratio equaling approximately 1 with stable reactivity and fuel isotopic composition. Nevertheless, to reach this state the reactor must still operate in an initial transient state for a lengthy period (over 10 years) of time, which requires implementing special measures concerning reactivity control.
The results obtained from calculations show the possibility of achieving a synergetic effect from combining two objectives. Using МА reprocessed from thermal reactor spent fuel in initial fuel loads in FR ensures a minimal reactivity margin during the entire fast reactor fuel operating period, comparable to the levels achieved in equilibrium state with any kind of relevant Pu isotopic composition. This should be combined with using reactivity compensators in the first fuel micro-campaigns.
In the paper presented are the results of simulation of the overall life cycle of a 1200 MWe fast reactor, reaching equilibrium fuel composition, and respective changes in spent fuel nuclide and isotopic composition. It is shown that МА from thermal and fast reactors spent fuel can be completely utilized in the new generation FRs without using special actinide burners.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Fast reactor
nitride fuel
spent nuclear fuel
minor actinides
reactivity margin
equilibrium cycle
Minor actinides transmutation in equilibrium cores of next generation FRs
Research Article
10.3897/nucet.5.39096
2019-12-10
nucet
Islamic Azad University, Tehran, Iran
author
Keshavarz, Sajad
https://orcid.org/0000-0003-1103-6810
Islamic Azad University, Tehran, Iran
author
Sardari, Dariush
2019-12-10
2019-12-10
2019
Nuclear Energy and Technology
2452-3038
5
361-371
2019
10.3897/nucet.5.39096
https://nucet.pensoft.net/article/39096/
https://nucet.pensoft.net/article/39096/download/pdf/
https://nucet.pensoft.net/article/39096/download/xml/
Gold nanoparticles can be used to increase the dose of the tumor due to its high atomic number as well as being free from apparent toxicity. The aim of this study is to evaluate the effect of distribution of gold nanoparticles models, as well as changes in nanoparticle sizes and spectrum of radiation energy along with the effects of nanoparticle penetration into surrounding tissues in dose enhancement factor DEF. Three mathematical models were considered for distribution of gold nanoparticles in the tumor, such as 1-uniform, 2- non-uniform distribution with no penetration margin and 3- non-uniform distribution with penetration margin of 2.7 mm of gold nanoparticles. For this purpose, a cube-shaped water phantom of 50 cm size in each side and a cube with 1 cm side placed at depth of 2 cm below the upper surface of the cubic phantom as the tumor was defined, and then 3 models of nanoparticle distribution were modeled. MCNPX code was used to simulate 3 distribution models. DEF was evaluated for sizes of 20, 25, 30, 50, 70, 90 and 100 nm of gold nanoparticles, and 50, 95, 250 keV and 4 MeV photon energies. In uniform distribution model the maximum DEF was observed at 100 nm and 50 keV being equal to 2.90, in non-uniform distribution with no penetration margin, the maximum DEF was measured at 100 nm and 50 keV being 1.69, and in non-uniform distribution with penetration margin of 2.7 mm, the maximum DEF was measured at 100 nm and 50 keV as 1.38, and the results have been showed that the dose was increased by injecting nanoparticles into the tumor. It is concluded that the highest DEF could be achieved in low energy photons and larger sizes of nanoparticles. Non-uniform distribution of gold nanoparticles can increase the dose and also decrease the DEF in comparison with the uniform distribution. The non-uniform distribution of nanoparticles with penetration margin showed a lower DEF than the non-uniform distribution without any margin and uniform distribution. Meanwhile, utilization of the real X-ray spectrum brought about a smaller DEF in comparison to mono-energetic X-ray photons.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
dose enhancement
gold nanoparticles
distribution
nanoparticle size
Monte Carlo
Different distributions of gold nanoparticles on the tumor and calculation of dose enhancement factor by Monte Carlo simulation
Research Article
10.3897/nucet.6.50863
2020-03-11
nucet
INM JSC PO Box 29, Zarechny, Russia
author
Portnykh, Irina
INM JSC PO Box 29, Zarechny, Russia
author
Kozlov, Aleksander
INM JSC PO Box 29, Zarechny, Russia
author
Panchenko, Valeriy
INM JSC PO Box 29, Zarechny, Russia
author
Barsanova, Svetlana
2020-03-11
2020-03-11
2020
Nuclear Energy and Technology
2452-3038
6
1
1-6
2020
10.3897/nucet.6.50863
https://nucet.pensoft.net/article/50863/
https://nucet.pensoft.net/article/50863/download/pdf/
https://nucet.pensoft.net/article/50863/download/xml/
The microstructures and physical properties of the austenitic Cr18Ni9-grade steel after 22 and 33 years of operation as part of the reactor internals were tested for assessing the conditions of the BN-600 reactor non-replaceable components (internals) and the potential of their subsequent use in predicting the reactor ultimate life. The paper presents histograms of the porosity distribution depending on the void size, in samples taken from portions that were subjected to neutron irradiation with displacement rates ranging from 1.0×10–9 to 4.3×10–8 dpa/s at temperatures from 370 to 440 °C. The elasticity characteristics were measured by resonance-type ultrasonic technique for the samples taken from the same portions of material. It was demonstrated that swelling calculated using the histograms of the porosity distribution depending on the void size has the maximum value at ~415 °C and after 33 years of irradiation reaches values of ~3%. Long-term variations of Young’s modulus demonstrate non-monotonous dependence on the damage dose. The maximum relative variation of Young’s modulus after 22 and 33 years of operation does not exceed 2% and 6%, respectively, of the values corresponding to the initial state. It was shown that along with the irradiation-induced swelling the changes in the physical properties are also affected in the process of irradiation by other structural changes and, in particular, by the formation of secondary phases. As shown by the results of the studies, operation of the BN-600 reactor internals made of Cr18Ni9-grade steel can be extended beyond 33 years of service. The comparison of the results obtained for the material after 22 and 33 years of operation contains information required for describing subsequent changes of the structure and properties of the Cr18Ni9 internals. The obtained results can be used for forecasting the reactor ultimate life within the framework of existing and developed models.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Internals
austenitic steel
neutron irradiation
porosity characteristics
irradiation-induced swelling
elasticity characteristics
A study into the structure and physical properties of the Cr18Ni9-grade steel following long-term irradiation as part of the BN-600 reactor internals
Research Article
10.3897/nucet.6.50865
2020-03-11
nucet
Obninsk Institute for Nuclear Power Engineering, Obninsk, Russia
author
Yuferov, Anatoliy
2020-03-11
2020-03-11
2020
Nuclear Energy and Technology
2452-3038
6
1
7-14
2020
10.3897/nucet.6.50865
https://nucet.pensoft.net/article/50865/
https://nucet.pensoft.net/article/50865/download/pdf/
https://nucet.pensoft.net/article/50865/download/xml/
A number of issues pertaining to comparative analysis of possible options for algorithmic and circuit embodiment of reactimeters were examined from the standpoint of the general theory of measuring instruments and the theory of digital filters. Structural diagrams of the linear part of the reactimeter, as well as the functional algorithms and their numerical implementation are described in terms of transient characteristics and transfer functions. Parallel, straight, canonical, symmetrized, lattice and ladder block structural diagrams are examined. The corresponding difference equations are given. The obtained results allow comparing possible circuit design solutions from the viewpoint of a number of criteria: the complexity of the elemental composition (the number of integrators, summation units, multipliers, delay elements), the number of necessary computing operations, the identifiability of the hardware function of the reactimeter, the coherence between the calculated and the measured values, the sensitivity to parameter uncertainties, etc. The possibility of considering the equations of the reactimeter as autoregressive is demonstrated, which ensures adaptability of the reactimeter under operating conditions. Certain algorithms for identification of the transient response characteristic and transfer function of the reactimeter are indicated. The possibility is shown of using identical algorithms in the main computing unit for solving the direct and inverse problems of nuclear reactor kinetics for ensuring consistency between the calculated and the measured reactivity values. Upper and lower estimated reactivity values are suggested for the moment of switching on the reactimeter. Implementation of such estimations in the reactimeter design allows minimizing the time needed for reaching by the reactimeter of the operating mode. Certain methodological simplifications were used in the development of ladder and lattice circuit design solutions. The database containing parameters of the instrumental functions of the circuit design solutions of the reactimeters is available on a public website. A number of tasks and directions for further research are identified.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Reactimeter
instrumental function
circuit design solution
variant analysis
Circuit design solutions for the reactimeters
Research Article
10.3897/nucet.6.50867
2020-03-11
nucet
National Research Nuclear University MEPhI, Moscow, Russia
author
Kulikov, Gennady
National Research Nuclear University MEPhI, Moscow, Russia
author
Shmelev, Anatoly
National Research Nuclear University MEPhI, Moscow, Russia
author
Apse, Vladimir
National Research Nuclear University MEPhI, Moscow, Russia
author
Kulikov, Yevgeny
2020-03-11
2020-03-11
2020
Nuclear Energy and Technology
2452-3038
6
1
15-21
2020
10.3897/nucet.6.50867
https://nucet.pensoft.net/article/50867/
https://nucet.pensoft.net/article/50867/download/pdf/
https://nucet.pensoft.net/article/50867/download/xml/
The purpose of the present study is the justification of the possibility of improving fast reactor safety by surrounding reactor cores with reflectors made of material with special neutron physics properties. Such properties of 208Pb lead isotope as heavy atomic weight, small neutron absorption cross section, and high inelastic scattering threshold result in certain peculiarities in neutron kinetics of the fast reactor equipped with 208Pb reflector, which can significantly enhance reactor safety. The reflector will also make possible generation of additional delayed neutrons characterized by the “dead” time. This will improve the resistibility of the fission chain reaction to stepwise reactivity excursions and exclude prompt supercriticality. Let us note that generation of additional delayed neutrons can be shaped by reactor designers. The relevance of the study amounts to the fact that generation of additional delayed neutrons in the reflector will make it possible mitigating the consequences of a reactivity accident even if the introduced reactivity exceeds the effective fraction of delayed neutrons. At the same time, the role of the fraction of delayed neutrons as the maximum permissible reactivity for reactor safety is depreciated. Scientific originality of the study pertains to the fact that the problem of yield of additional neutrons with properties close to normal delayed neutrons, has not been posed before. The authors suggest a new method for enhancing safety of fast reactors by increasing the fraction of delayed neutrons due to the time delay of prompt neutrons during their transfer in the reflector. In order to benefit from the expected advantages, the following combination is acceptable: lead enriched by 208Pb is used as a neutron reflector while natural lead or other material (sodium, etc.) is used as a coolant in the reactor core.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Nuclear safety
reactivity accident
delayed neutrons
fast reactor
radiogenic lead
Safety features of fast reactor with heavy atomic weight weakly neutron absorbing reflector
Research Article
10.3897/nucet.6.50868
2020-03-11
nucet
Obninsk Institute for Nuclear Power Engineering, Obninsk, Russia
author
Khorasanov, Georgiy
Nuclear Safety Institute of the RAS, Moscow, Russia
author
Blokhin, Anatoliy
2020-03-11
2020-03-11
2020
Nuclear Energy and Technology
2452-3038
6
1
23-27
2020
10.3897/nucet.6.50868
https://nucet.pensoft.net/article/50868/
https://nucet.pensoft.net/article/50868/download/pdf/
https://nucet.pensoft.net/article/50868/download/xml/
The paper considers the concept of a fast lead cooled 25MW reactor for a variety of applications, including incineration of minor actinides, production of medical radioisotopes, testing of radiation-damaged nuclear technology materials, etc. A specific feature of the proposed reactor is rather a high neutron flux of 2.6·1015 n/(cm2·s) at the core center, high average neutron energy of 0.95 MeV at the core center, and a large fraction (40%) of hard neutrons (En > 0.8 MeV). The extremely high estimated reactor parameters are achieved thanks to the small core dimensions (DxH ≈ 0.50×0.42 m2), innovative metallic fuel of the Pu-Am-Np-Zr alloy, and the 208Pb enriched lead coolant. A relatively high probability of 241Am fission (about 50%) is achieved in the reactor core’s hard spectrum, this making it possible to incinerate up to 4 kg of 241Am during one reactor campaign of 1000 effective days.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Small reactor
plutonium-americium-neptunium fuel
lead-208 coolant
incineration of minor actinides
Reactor with metallic fuel and lead-208 coolant
Research Article
10.3897/nucet.6.51252
2020-03-27
nucet
“KVANT PROGRAMM” LTD, Moscow, Russia
author
Maksimov, Ivan
Bauman Moscow State Technical University, Moscow, Russia
author
Vladimir, Vladimir
2020-03-27
2020-03-27
2020
Nuclear Energy and Technology
2452-3038
6
1
29-35
2020
10.1016/j.isatra.2010.08.004
10.1007/BF01386390
10.1016/j.pnucene.2005.03.008
2009
Instrumentation important to safety.
2009
75 pp
IEC 60988 Nuclear power plants (2009) Instrumentation important to safety.Acoustic monitoring systems for detection of loose parts: Characteristics, design criteria and operational procedures. International Electrotechnical Commission, 75 pp.
2015
2015
ISO 13379-1-2015 (2015) Condition monitoring and diagnostics of machines – Data interpretation and diagnostics techniques – Part 1: General guidelines. Moscow. Standartinform Publ., 33 pp. [in Russian]
HI
Ki
author
2017
ANN Based Localization of Metal Ball Impacts on Reactor Pressure Boundary Structure. Transactions of the Korean Nuclear Society.
2017
3 pp
10.1016/j.sigpro.2005.01.012
10.1016/j.pnucene.2017.05.004
10.1016/j.ymssp.2016.10.003
10.18698/0236-3933-2018-1-4-15
10.1016/0149-1970(85)90086-1
2015
2015
Operation and Maintenance of Nuclear Power Plants (2015) Part 12. Loose Part Monitoring. American Society of Mechanical Engineers (ASME), 523 pp.
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10.1088/0957-0233/17/10/030
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1981
Regulatory Guide 1.133 (1981) Loose Part Detection Program for the Primary System of Light-Water Cooled Reactors: tech. rep. U.S. Nuclear Regulatory Commission, 6 pp.
10.1016/S0029-5493(02)00281-9
10.1016/S0149-1970(98)00004-3
C
Truong
author
2018
2018
10.1016/j.pnucene.2005.09.012
10.1117/12.819732
10.3897/nucet.6.51252
https://nucet.pensoft.net/article/51252/
https://nucet.pensoft.net/article/51252/download/pdf/
https://nucet.pensoft.net/article/51252/download/xml/
As operational experience shows, it can hardly be excluded that some detached or loosened parts and even foreign objects (hereinafter referred to as the ‘loose parts’) may appear in the main circulation loop of VVER reactor plants. Naturally, the sooner such incidents are detected and evaluated, the more time will be available to eliminate or at least minimize damage to the reactor plant main equipment. The paper describes a method for localizing the impact of loose parts located in the coolant circulation circuit of a VVER reactor plant. To diagnose malfunctions of the reactor plant main equipment, it is necessary to accurately determine the place where the acoustic anomaly occurred. Therefore, if some loose parts make themselves felt, it is important to track the path of their movement along the main circulation circuit as well as their location using physical barriers. The method is based on the representation of the surface, along which an acoustic wave travels, as a 3D model of the reactor plant (RP) main circulation circuit. The model has the form of a graph in which the vertices characterize the control points on the RP surface and the edges are the distances between them. The method uses information about the acoustic wave velocity and the time difference of arrivals (TDOAs) of the signal received by various sensors. It is shown that, when the effect is received by more than three sensors, along with an estimate of the impact coordinate, it becomes possible to estimate the average acoustic wave velocity. To determine time of arrival, the signal dispersion change point detection method is used. Provided that the average size between the control points on the RP surface was 300 mm, the average localization error was about 600 mm. The developed algorithm can be easily adapted to any VVER reactor plant. The obtained deviation values are acceptable for practical use.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
Loose Parts Monitoring System (LPMS)
diagnostics of NPPs
impact localization
acoustic wave
main circulation circuit
VVER reactor plant
A localization method for loose parts monitoring system of VVER reactor plants
Research Article
10.3897/nucet.6.51253
2020-03-27
nucet
Sosny R&D Company, Dimitrovgrad, Russia
DETI MEPhI, Dimitrovgrad, Russia
author
Shamsutdinov, Rinat
Sosny R&D Company, Dimitrovgrad, Russia
DETI MEPhI, Dimitrovgrad, Russia
author
Pavlov, Sergey
Sosny R&D Company, Dimitrovgrad, Russia
author
Leshchenko, Anton
2020-03-27
2020-03-27
2020
Nuclear Energy and Technology
2452-3038
6
1
37-42
2020
AA
Alyamovsky
author
2005
2005
10.1007/s11630-015-0756-4
D
Bestion
author
2014
2014
MA
Bykov
author
2018
2018
10.1016/0029-5493(68)90086-1
C
Hirsch
author
1991
1991
2002
2002
IAEA-TECDOC-1379 (2002) Use of computational fluid dynamics codes for safety analysis of nuclear reactor systems: 1–50.
J
Joshi
author
2019
Advances of Computational Fluid Dynamics in Nuclear Reactor Design and Safety Assessment.
2019
888 pp
J
Mahaffy
author
2015
2015
K
Mizzi
author
2015
2015
SV
Patankar
author
1980
1980
10.2516/ogst/2015019
PJ
Roache
author
1998
Fundamentals of Computational Fluid Dynamics.
1998
648 pp
Y
Saad
author
1996
1996
10.1007/s10512-019-00488-3
BL
Smith
author
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2015
A
Sobachkin
author
2014
2014
10.1017/aer.2015.10
10.1007/s00158-016-1643-7
10.1007/s11663-015-0342-x
DC
Wilcox
author
1998
1998
10.3897/nucet.6.51253
https://nucet.pensoft.net/article/51253/
https://nucet.pensoft.net/article/51253/download/pdf/
https://nucet.pensoft.net/article/51253/download/xml/
The retort batch furnace for the carbothermal synthesis of uranium and plutonium nitrides is a component of the mixed nitride uranium-plutonium (MNUP) fuel Fabrication/Refabrication Module (FRM), a part of the Pilot Demonstration Energy Complex (PDEC). A CFD model of the furnace was built in the SolidWorks Flow Simulation code to check the feasibility of its thermophysical operating modes (heating and cooling rates, temperature of the product loaded into the furnace). The experimental data obtained from the performance tests was used to verify the developed CFD model and to confirm its adequacy. The relative deviation of the calculated temperature of the loaded product from the experimental data in the process of isothermal annealing does not exceed 0.7%. The temperatures of the loaded product predicted by the CFD model were used to justify engineering solutions for the uranium and plutonium nitrides carbothermal synthesis furnace. The CFD model can be used to define the furnace operation mode by selecting the gas flow rates inside and outside the retort, the heater temperature, and heating and cooling rates for the product loaded in the furnace.
text/html
en_US
National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
CFD model
MNUP fuel
SolidWorks Flow Simulationv
furnace
carbothermal synthesis
mathematical model
grid model
Development and verification of CFD model of carbothermal synthesis furnace for production of mixed nitride uranium-plutonium fuel
Research Article
10.3897/nucet.6.51778
2020-03-27
nucet
Obninsk Institute for Nuclear Power Engineering, NRNU “MEPhI”, Obninsk, Russia
author
Sobolev, Artem
Obninsk Institute for Nuclear Power Engineering, NRNU “MEPhI”, Obninsk, Russia
author
Danilov, Pavel
2020-03-27
2020-03-27
2020
Nuclear Energy and Technology
2452-3038
6
1
43-47
2020
AM
Amelin
author
2010
2010
AM
Amelin
author
2012
2012
1988
1988
GOST 25461-82 (1988) Transport packages for spent fuel assemblies of nuclear reactors. Requirements for methods of calculating nuclear safety. https://meganorm.ru/Data/77/7743.pdf [accessed June 17, 2019] [in Russian]
1984
1984
GOST 26013-83 (1984) Transport packages for spent fuel assemblies of nuclear reactors. General technical requirements. https://meganorm.ru/Index2/1/4294828/4294828423.htm [accessed June 17, 2019] [in Russian]
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2014
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GOST 5632-72 (2015) High-alloy steels and corrosion-resistant, heat-resistant alloys. https://meganorm.ru/Data/421/42169.pdf [accessed June 17, 2019] [in Russian]
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NG
Gusev
author
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Guskov
author
2012
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IAEA SSR-6 (2012) Regulations for the Safe Transport of Radioactive Material. https://www-pub.iaea.org/MTCD/Publications/PDF/P1570r_web.pdf [accessed June 17, 2019] [in Russian]
VP
Mashkovich
author
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NP-053-16 (1995) Federal norms and rules in the field of the use of atomic energy “Safety rules for the transport of radioactive materials”. https://meganorm.ru/Data2/1/4293748/4293748284.htm [accessed June 17, 2019] [in Russian]
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Opalovsky
author
2007
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Opalovsky
author
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2013
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PBYA-06-00-96 (2013) Basic industry nuclear safety rules for the use, processing, storage and transportation of nuclear hazardous fissile materials. https://meganorm.ru/Data2/1/4293842/4293842541.htm [accessed June 17, 2019] [in Russian]
2016
Basic nuclear safety rules for the production, use, processing, storage and transportation of nuclear fissionable materials
2016
PBYA-06-09-2016 (2016) Basic nuclear safety rules for the production, use, processing, storage and transportation of nuclear fissionable materials. https://meganorm.ru/Data2/1/4293746/4293746881.pdf[accessed June 17, 2019]. [in Russian]
AS
Vasilyev
author
2019
2019
VV
Vorontsov
author
2008
2008
10.3897/nucet.6.51778
https://nucet.pensoft.net/article/51778/
https://nucet.pensoft.net/article/51778/download/pdf/
https://nucet.pensoft.net/article/51778/download/xml/
The paper discusses the stages of calculating the radiation safety of spent nuclear fuel (SNF) transport packages, in particular, transport casks and some related problems. The problem of describing the source of neutrons and gamma radiation of spent nuclear fuel is shown. For individual designs of fuel assemblies, data are given on isotopes that make the main contribution to the neutron source as well as on gamma rays in nuclear fuel material and structural materials. The authors emphasize the necessity of analyzing the influence of the initial spent fuel parameters on the formation of the radiation spectrum and, therefore, on the radiation situation around the transport casks. Consideration is given to the problem of assessing the attenuation of gamma radiation in calculating protection analytically and using software. Due to the ambiguity of the position of the zone with the highest effective dose value on the SNF transport cask surface, it is indicated that preliminary estimates are required to take into account all radiation sources and their nonuniformities. All the problems presented in the paper are currently being solved by means of rather complex and voluminous calculations that take a long time. In order to be able to conduct a preliminary assessment of the radiation situation around the transport casks, the authors propose to create a methodology that will determine the type of interrelations between the maximum effective dose and input parameters, such as fuel burnup, decay, fuel composition, protection material in the SNF transport cask, etc. This methodology will make it possible to improve the efficiency of the process of designing the SNF transport casks, avoid possible design errors and, in particular, when used as intended, resolve the issue of the SNF cask loading configuration.
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National Research Nuclear University MEPhI (Moscow Engineering Physics Institute)
SNF transport cask
radiation safety
neutron source
gamma source
radiation protection calculation
Problems of radiation safety calculations related to spent nuclear fuel transport casks
Research Article
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