Research Article |
Corresponding author: Sergey V. Bedenko ( bedenko@tpu.ru ) Academic editor: Yury Kazansky
© 2022 Sergey V. Bedenko, Igor O. Lutsik, Anton A. Matyushin , Sergey D. Polozkov , Vladimir M. Shmakov, Dmitry G. Modestov, Vadim V. Prikhodko , Andrey V. Arzhannikov .
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:
Bedenko SV, Lutsik IO, Matyushin AA, Polozkov SD, Shmakov VM, Modestov DG, Prikhodko VV, Arzhannikov AV (2022) Fusion-fission hybrid reactor facility: neutronic research. Nuclear Energy and Technology 8(1): 25-30. https://doi.org/10.3897/nucet.8.82294
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The authors investigate the neutronic characteristics of the operating mode of a hybrid nuclear-thermonuclear reactor. The facility under study consists of a modified core of a high-temperature gas-cooled thorium reactor and an extended plasma neutron source penetrating the near-axial region of the core. The proposed facility has a generated power that is convenient for the regional level (60–100 MW), acceptable geometric dimensions and a low level of radioactive waste.
The paper demonstrates optimization neutronic studies, the purpose of which is to level the resulting offsets of the radial energy release field, which are formed within the fuel part of the blanket during long-term operation and due to the pulsed operation of the plasma D-T neutron source.
The calculations were performed using both previously developed models and the SERPENT 2.1.31 precision program code based on the Monte Carlo method. In the simulation, we used pointwise evaluated nuclear data converted from the ENDF-B/VII.1 library, as well as additional data for neutron scattering in graphite from ENDF-B/VII.0, based on the S (α, β) formalism.
Fusion-fission hybrid reactor facility, plasma D-T-neutron generator, neutronic research
Thermonuclear research is being conducted on an interstate scale and is aimed at the prospect of entering industrial energy production after 2050. On this path, in 2020, researchers at the Korea Institute of Fusion Energy (KFE) managed to achieve plasma confinement in a toroidal magnetic trap (KSTAR tokamak (
Power production plants using tokamaks will be of exceptionally large size and capacity; they will be built in the distant future. Our research is focused on the prospect for the practical use of thermonuclear power in a shorter period (
The reactor facility under study is a hybrid reactor, the core (blanket) of which consists of an assembly of unified fuel elements of a high-temperature gas-cooled thorium reactor (HGTRU, Tomsk Polytechnic University, Tomsk (
The paper describes optimization neutronic studies, the purpose of which is to level the resulting offsets of the radial energy release field, which are formed within the fuel part of the blanket during long-term operation and due to the pulsed operation of the plasma D-T neutron source. The calculations were performed using the SERPENT 2.1.31 precision program code based on the Monte Carlo method (
The reactor facility under study (Fig.
Scheme of the design solution for the fusion-fission hybrid reactor facility a) 3D design model of the facility, including the blanket with Th-Pu fuel and the extended PNS (1–4 are the numbers of rows with fuel and non-fuel elements); b) fuel element of a unified design (
The PNS (a scheme of the computational model is in Fig.
The plasma pinch zone (Fig.
It should be noted that in the considered ‘plasma neutron source – subcritical blanket’ configuration, the high-temperature plasma pinch is formed in a repetitively pulsed mode and propagates from the near-axial region over the entire multiplying part in time correlation with the PNS.
The simulation results (Fig.
At the periphery of the fuel part (Fig.
Note that reaching the steady state in the entire multiplying region is observed in the time interval from 100 ms to 1 s. In a second, the total number of fissions in the blanket increases to 20 (per one neutron coming from the PNS into the blanket in the radial direction). This value does not change any further, providing heating of the blanket at a rate of no more than 10 K×h–1 with a constant neutron emission from the PNS at a level of 5.76×1017 n×s–1, which meets the requirements of thermal engineering reliability during a cold startup.
Neutronic calculations were performed using the SERPENT 2.1.31 precision program code based on the Monte Carlo method (
The neutronic optimization of the facility was performed by profiling the power density along the radius of the fuel part of the blanket by changing the content of the Pu fraction. In order to reduce the power density in the near-axial region, the first row of the fuel elements, even before profiling, was immediately replaced by graphite elements with holes for the helium coolant (Fig.
As expected, the most power-intensive part is in the first row adjacent to the PNS. The calculation showed that the maximum peak power density of an unprofiled blanket reaches a level of 1.25% (Fig.
After profiling (Fig.
The subcriticality value required for such systems is achieved through the use of burnable poisons (BP). Table
The burnable poison was used in two placement options (Table
The options 01 and 02 are non-profiled and profiled fuel parts of the facility blanket, respectively. The option 03_ZrB2 is a heterogeneous way of placing the burnable absorber, representing a technological solution proposed in (
An analysis of the results given in Tab. showed that the best reactivity compensation options, in terms of neutronic characteristics, are 03_ZrB2 and 07_Er2O3Hom (Table
In further calculations, the 03_ZrB2 option was used, since the technology for applying such coatings was developed at the Tomsk Polytechnic University (
Note that the system of rods intended for control and emergency protection is not calculated in the configuration under study, since the facility is in a subcritical state throughout the entire operating cycle (Fig.
It should also be noted that the use of PNS as an additional neutron source increases the nuclear safety of the facility, since when the injection of neutral atoms is turned off, the neutron generation drops by about a factor of two over the first 2.5 ms and another 20 times over the next 5 ms (Fig.
This result indicates that the decrease in the generation of additional neutrons in the blanket proceeds much faster than it occurs in the core of a conventional reactor.
Calculation results for various reactivity compensation options of the facility blanket
Calculation option | H. met. mass, kg | Pu mass, kg | BP mass, kg | Exposure time (250 MW×day/kg), g |
---|---|---|---|---|
01_non-profiled blanket (Core) | 290.77 | 147.57 | – | 3.32 |
02_profiled blanket (Core) | 302.33 | 153.44 | – | 3.45 |
03_ZrB2 | 305.05 | 154.82 | 5.23 | 3.48 |
04_Gd2O3Hom | 277.12 | 157.76 | 22.86 | 3.16 |
07_Er2O3Hom | 238.34 | 160.48 | 56.18 | 2.72 |
09_HfO2Hom | 235.96 | 157.85 | 57.73 | 2.69 |
11_Pa-231Hom | 233.78 | 157.41 | 76.04 | 3.53 |
The paper describes the neutronic and thermophysical optimization of the operating mode of the facility:
The reactivity control system and the use of permanent reactivity compensators for this blanket configuration is not provided, since the facility is in a subcritical state (calculation options 03_ZrB2, 07_Er2O3Hom, 11_Pa-231Hom) throughout the entire operating cycle, and all control of the operating mode is carried out by varying the neutron flux coming from the PNS.
The work was supported by the Russian Foundation for Basic Research (Grant No. 19-29-02005).