Corresponding author: Vadim Naumov ( naumovva@goi.kolasc.net.ru ) Academic editor: Yury Korovin
© 2018 Vadim Naumov, Sergey Gusak, Andrey Naumov.
This is an open access article distributed under the terms of the Creative Commons Attribution License (CC BY 4.0), which permits unrestricted use, distribution, and reproduction in any medium, provided the original author and source are credited.
Citation:
Naumov V, Gusak S, Naumov A (2018) Small nuclear power plants for power supply in arctic regions: assessment of spent nuclear fuel radioactivity. Nuclear Energy and Technology 4(2): 119-125. https://doi.org/10.3897/nucet.4.30677
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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.
Arctic regions of Russia, small nuclear power plants, reactors, spent nuclear fuel, fuel cycle, radioactivity
The need to develop alternative energy sources and to implement of upgrades of power generation infrastructure in Arctic regions was included among the priority tasks implementation of which is aimed at the achievement of main purposes of state policy of the Russian Federation in the Arctic regions (
The study is dedicated to the assessment of radioactivity of SNF which was performed on the basis of mathematical simulation of fuel cycles of SNPP reactors of different types and of prototype reactor facilities. Radionuclides producing main contribution in SNF radioactivity during the stages of irradiated fuel handling after its cooling down in reactor SNF storage facilities including the following: β-active 85Kr (T1/2 = 10.9 years), 90Sr (T1/2 = 28.6 years), 137Cs (T1/2 = 30.1 years), 151Sm (T1/2 = 90 years) and α-active 238,239,240,241,242Pu, 236U, 237Np, 241Am (T1/2 = 433 years), 244Cm (T1/2 = 18.1 years) were examined in the studies and in the analysis of results.
Currently a number of research institutes (Dollezhal Research and Design Institute of Energy Technologies (NIKIET), IPPE) and design bureaus (Afrikantov Experimental Design Bureau for Mechanical Engineering (OKBM), OKB Gidropress) developed on the basis of existing experience of nuclear-powered shipbuilding several options of design of shipboard nuclear propulsion facilities of different types and configurations which can be used for covering prospective power demand from potential consumers in Arctic regions of the RF (
The following two reactor facilities are regarded as prototype facilities for the class of light water reactors: shipboard reactor facility of OK-900A type on “Siberia” nuclear-powered icebreaker which was operated from 1978 to 1992 and generated 84 GW·day of thermal power during the first Arctic navigation (1978 – 1980), as well as reactor facility of KLT-40 type used to power “Sevmorput” nuclear lighter carrier. Two cores of KLT-40 reactor were spent during the period from 1988 to 1999 with average power output equal to 78 GW·day (
Simplified mathematical models of reactor cores (RC) of the reactors under examination with description of neutronics processes using KRATER software complex (SW) (
FC parameters and characteristics of cores of reactors under study used as the input data for constructing mathematical models are presented in Tables
Parameters of fuel cycles of SNPP reactor installations and their prototypes.
SNPP or shipboard reactor facility | Installed thermal power, MW | Power generation capacity, GW·day | Time of installed power operation, years | CF | Fuel cycle duration, years | Time between core refueling, years |
ABV | 45 | 131.5 | 8.0 | 0.8 | 10 | 10–12 |
UNITERM | 30 | 181 | 16.5 | 0.8 | 20.6 | 25 |
KLT-40S | 150 | 137.5 | 2.51 | 0.65 | 3.9 | 4 |
RITM-200M | 175 | 291.7 | 4.57 | 0.65 | 7.03 | 10-12 |
KLT-40, “Sevmorput” nuclear lighter carrier | 135 | 78 | 1.58 | ~0.3 | ~5.5 | ~6 |
OK-900A, “Siberia” nuclear icebreaker* | 171 | 84 | 1.35 | ~0.6 | 2.3 | 4 |
SVBR-100 | 280 | 631 | 6.18 | 0.9 | 6.9 | 8 |
SVBR-10 | 43.3 | 243 | 15.4 | 0.8 | 19 | 20–21 |
NSM, Project 705K | 150 | 25 | 0.46 | ~0.09 | ~5.0 | – |
Characteristics of SNPP reactors and their prototypes.
SNPP or shipboard reactor facility | Uranium load, t | Mass of fuel composition, t | Average uranium enrichment with 235U isotope, % | Fuel composition | Average fuel burnup depth, g/cm3 |
ABV | 1.4 | 1.9 | 16.5 | UO2 in aluminum-silicon matrix | 0.43 |
UNITERM | 1.58 | 2.52 | 19.7 | UO2 in zirconium matrix | 0.665 |
KLT-40S | 1.53 | 2.09 | 17.4 | UO2 in aluminum-silicon matrix | 0.429 |
RITM-200M | 3.2 | 4.28 | 17.5 | UO2 in aluminum-silicon matrix | 0.429 |
KLT-40, “Sevmorput” nuclear lighter carrier | 0.167 | 0.84 | 90 | Uranium-zirconium alloy | 0.35 |
OK-900A, “Siberia” nuclear icebreaker* | 0.513 | 0.95 | 40.6 | UAl3 in aluminum matrix | 0.38 |
SVBR-100 | 9.188 | 10.4 | 16.5 | UO2 | 0.62 |
SVBR-10 | 4.037 | 4.58 | 18.7 | UO2 | 0.62 |
NSM, Project 705K | 0.182 | 0.4 | 89 | UBe13 in beryllium matrix | < 0.1 |
Characteristics of cores of SNPP reactors and their prototypes.
SNPP or shipboard reactor facility | Number of fuel assemblies (FA grid pitch, cm) | Fuel rod diameter, mm (fuel cladding material) | Number of fuel rods in the core | Reactor core diameter/height, m |
ABV | 121 (10) | 6.8×0.5 (alloy Э-110) | 9317 | 1.155/1.3 |
UNITERM | 265 (7.2) | 5.8×0.5 (alloy Э-110) | 14310 | 1.231/1.1 |
KLT-40S | 121 (10) | 6.2×0.5 (alloy Э-635) | 12342 | 1.155/1.3 |
RITM-200M | 199 (10) | 6.2×0.5 (steel) | 20467 | 1.48/1.65 |
KLT-40, “Sevmorput” nuclear lighter carrier | 241 (7.2) | 5.8×0.5 (alloy Э-110) | 12787 | 1.155/1.0 |
OK-900A, “Siberia” nuclear icebreaker* | 241 (7.2) | 5.8×0.5 (steel) | 12787 | 1.155/1.0 |
SVBR-100 | 61 (20) | 12×0.4 (steel) | 12078 | 1.646/0.9 |
SVBR-10 | 27 (20)* | 12×0.4 (steel) | 5373 | 1.086/0.9 |
NSM, Project 705K | ** | 11×0.5 (steel) | 4200 | 0.885/0.928 |
Heterogenous core of channel type of reactor facility OK-900A consists of 241 pressure channels (PC) each of which represents Ø60×1 mm pipe made of zirconium-niobium alloy containing fuel rod bundle consisting of 61 rods (54 fuel rods and seven absorber rods (AR)). Cross-section of pressure channel is shown in Fig.
Cross section of pressure channel of light water reactor core: a) pressure channel 10-14-3М of icebreaker OK-900A reactor core У (
Mathematical model of OK-900A reactor facility was developed on the basis of the data in Tables
Comparison of masses and activities of long-lived radionuclides in reactor core by the end of fuel cycle of light-water SNPP reactors and their prototypes * (KRATER software).
Parameter | Reactor facilities | |||||
KLT-40, ”Sevmorput” nuclear lighter carrier | OK-900A, “Siberia” nuclear icebreaker | UNITERM | ABV | KLT-40S | RITM-200M | |
Mass of 235U, kg | 51.7 | 100 | 108 | 88 | 119 | 243 |
Mass of 237Np, kg | 1.01 | 1.12 | 3.34 | 2.02 | 2.47 | 5.38 |
Mass of 238Pu, kg | 0.345 | 0.21 | 0.82 | 0.526 | 0.606 | 1.32 |
Mass of 238U, kg | 14.9 | 296 | 1220 | 1128 | 1224 | 2544 |
Mass of 239Pu, kg | 0.41 | 4.21 | 12.4 | 9.77 | 13.3 | 28.7 |
Mass of 240Pu, kg | 0.16 | 1.11 | 4.39 | 3.49 | 3.79 | 8.24 |
Mass of 241Pu, kg | 0.13 | 0.92 | 3.40 | 2.57 | 3.22 | 7.15 |
Mass of 241Am, kg | 0.004 | 0.043 | 0.87 | 0.31 | 0.156 | 0.636 |
Mass of 244Cm, kg | 0.001 | 0.01 | 0.036 | 0.025 | 0.027 | 0.0575 |
Mass of 85Kr, kg | 0.074 | 0.079 | 0.098 | 0.093 | 0.116 | 0.221 |
Mass of 90Sr, kg | 1.68 | 1.825 | 3.03 | 2.40 | 2.68 | 5.45 |
Mass of 137Cs, kg | 2.78 | 3.12 | 5.66 | 4.42 | 4.91 | 10.1 |
Mass of 151Sm, kg | 0.01 | 0.026 | 0.036 | 0.028 | 0.041 | 0.087 |
Total a-activity, PBq | 0.222 | 0.19 | 0.817 | 0.50 | 0.548 | 1.225 |
Specific a-activity, TBq/kg | 0.26 | 0.20 | 0.324 | 0.263 | 0.265 | 0.286 |
Total b-activity**, PBq | 35.7 | 39.4 | 68 | 53.6 | 60 | 122 |
Specific b-activity**, TBq/kg | 42.5 | 41.6 | 27 | 28.3 | 28.9 | 28.5 |
UNITERM SNPP reactor is the closest to the investigated prototype with regard to reactor core design. This rector has design of pressure channels similar to OK-900A reactor (Fig.
The next group of reactor facilities under investigation with cognate accepted design and technological solutions with regard to reactor core is formed by ABV, KLT-40S and RITM-200M. Fuel cycle of these reactor facilities differs from the prototype ones by large values of power generation capacity and time of installed power operation (see Table
Cross-section of fuel assembly of cassette-type core of SNPP reactors: а) cassette of ABV-6 reactor (IAEA-TECDOC-1536); b) Cassette of KLT-40S reactor (
Heterogenous structure of cassette-type reactor cores is accounted for in mathematical models of fuel cycles using five-zone elementary cylindrical reactor cell (Fig.
Two projects of fast reactors with lead-bismuth coolant SVBR-10 and SVBR-100 intended for application as power sources in remote regions of Russia were developed in our country. Fuel cycle of SVBR-100 reactor was investigated in (
Comparison of masses and activities of long-lived radionuclides in the reactor core by the end of fuel cycle of liquid metal SNPP fast reactors and of their prototype – reactor of Project 705K NSM (masses of 236U and 242Pu are not presented).
Parameter | Intermediate neutron reactor of Project 705K NSM (KRATER SW) | Fast SNPP reactors | |
SVBR-10 (KRATER SW) |
SVBR-100 (KRATER SW and ( |
||
Mass235U, kg | 126.4 | 504 | 941 |
Mass237Np, kg | 0.87 | 3.0 | 6.77 |
Mass238Pu, kg | 0.082 | 0.31 | 0.814 |
Mass238U, kg | 16.48 | 3113 | 7220 |
Mass239Pu, kg | 1.33 | 111 | 331 |
Mass240Pu, kg | 0.086 | 4.75 | 16.4 |
Mass241Pu, kg | 0.062 | 0.175 | 0.53 |
Mass241Am, kg | 0.001 | 0.013 | 0.0368 |
Mass244Cm, kg | 0.000003 | 0.00001 | 0.000022 |
Mass85Kr, kg | 0.0252 | 0.143 | 0.472 |
Mass90Sr, kg | 0.559 | 4.11 | 11.98 |
Mass137Cs, kg | 0.917 | 7.61 | 21.83 |
Mass151Sm, kg | 0.042 | 0.685 | 1.71 |
Total a-activity, PBq | 0.056 | 0.597 | 1.49 |
Specific a-activity, TBq/kg | 0.142 | 0.13 | 0.143 |
Total b-activity, PBq | 11.9 | 92.1 | 263 |
Specific b-activity, TBq/kg | 30 | 20 | 25 |
Structure of reactor core and its radial reflector is shown in Fig.
Mode of operation at nominal thermal power with generation of 243 GW·day and time of installed power operation equal to 15.4 years was examined for calculating isotopic composition of SVBR-10 reactor core. Rector core is configured using the same fuel assemblies as in SVBR-100 reactor. Number of FAs was taken to be equal to 27.
Results of investigation of isotopic composition and activity of SNF by the end of fuel cycle of SVBR-10 and SVBR-100 reactors are presented in Table
Main results of the study provided in Tables
The authors regard the comparison of total specific activities of radiation dose shaping FPs within each class of SNPP reactor facilities to be of utmost scientific and practical interest, because this comparison allows formulating judgement about possible differences in radiation conditions of handling SNF of SNPP reactor facilities and their prototypes. In case of light water reactor types specific activities of long-lived fission products in SNF of all SNPP reactors have the values of activity of about 27 TBq/kg, while those for prototype reactor facilities are equal to ~42 TBq/kg. Thus, it is evident that intensity of sources of ionizing radiation emitted in decays of 85Kr, 137Cs and 137mBa is by approximately 1.5 lower for SNF of SNPP reactor facilities compared to their prototypes. It follows from the above that handling SNF of SNPP reactor facilities will be performed at time moments after cooling SNF down in reactor spent fuel storage facilities at lower levels of ionizing radiation than those for SNF of prototype reactor facilities. In our opinion this fact allows formulating the conclusion on the possibility (from the viewpoint of ensuring radiation safety) of application of the SNF handling infrastructure used at present on prototype reactor facilities with SNF handling technology for SNPP reactor facilities currently under development.
In case of liquid-metal fast reactors specific β-activity (and, consequently, gamma-activity of 137Cs→137mBa decay chain as well) of SNF of reactor facility of SVBR type is also lower than that for the prototype (reactor facility of Project 705K NSM) by 1.2 – 1.5 times. Conclusion similar to that for SNPP reactor facility with light-water reactors can be formulated in this case.
As to the analysis of mass composition of actinoids in SNF of light-water SNPP reactors, significant accumulation of plutonium isotopes in SNF (from 7 kg per ton of uranium in the case of ABV reactor to 9 kg/t for RITM-200M) has to be noted, which is explained by high (about 2%) burnup of 238U because of high values of power yield by cores of reactor facilities of this class. This result is rather hard to predict, if the fact is taken into account of a lower specific content of 239Pu in spent nuclear fuel of commercial reactors of VVER-440 and VVER-1000 types, where less enriched fuel is used compared to SNPP reactors. The determined high concentration of 239Pu in SNF of SNPP reactor facility with VVER type reactors allows formulating the conclusion on the advisability of radiochemical reprocessing of SNF of SNPP reactor facilities.
Mathematical models were developed of neutronics processes in reactor cores of SNPP of the following two classes: on the basis of light-water reactors and liquid-metal fast reactors, as well as of prototype reactor facilities. Calculation study was performed of accumulation of long-lived radiation dose shaping fission products and actinoids in SNF of these reactors.
Analysis of specific activity of long-lived radiation dose shaping fission products (85Kr, 90Sr, 137Cs) established that this parameter characterizing intensity of sources of ionizing radiation is lower for SNF reactors of SNPP than for SNF of prototype reactors, which allows forecasting the possibility, in accordance with conditions for ensuring radiation safety, of application of SNF handling infrastructure currently used for prototype reactors for handling SNF of SNPP reactors at time moments after extraction of SNF after cooling down in on-site reactor spent fuel storage facilities.
Significant accumulation of 239Pu in SNF of light-water SNPP reactors (7 – 9 kg per ton uranium) was discovered based on the analysis of mass composition of long-lived actinoids, which exceeds concentration of 239Pu equal to ~ 5.5 kg per ton uranium in SNF of commercial reactors of VVER-440 and VVER-1000 types, which is the indication of the advisability of reprocessing SNF of SNPP reactors.