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Title: Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

Abstract

Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the {sup 231}Pa–{sup 232}U–{sup 233}U–Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of {sup 232}U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

Authors:
; ; ; ; ;  [1]
  1. National Research Nuclear University MEPhI (Moscow Engineering Physics Institute) (Russian Federation)
Publication Date:
OSTI Identifier:
22471976
Resource Type:
Journal Article
Resource Relation:
Journal Name: Physics of Atomic Nuclei; Journal Volume: 78; Journal Issue: 10; Other Information: Copyright (c) 2015 Pleiades Publishing, Ltd.; Country of input: International Atomic Energy Agency (IAEA)
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; BREEDING BLANKETS; BURNUP; FISSILE MATERIALS; FISSION; FUEL CYCLE; NUCLEAR FUELS; PROTACTINIUM 231; THERMONUCLEAR REACTORS; THORIUM; URANIUM; URANIUM 232; URANIUM 233; WATER COOLED REACTORS; WATER MODERATED REACTORS

Citation Formats

Shmelev, A. N., E-mail: shmelan@mail.ru, Kulikov, G. G., E-mail: ggkulikov@mephi.ru, Kurnaev, V. A., E-mail: kurnaev@yandex.ru, Salahutdinov, G. H., E-mail: saip07@mail.ru, Kulikov, E. G., E-mail: egkulikov@mephi.ru, and Apse, V. A., E-mail: apseva@mail.ru. Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors. United States: N. p., 2015. Web. doi:10.1134/S1063778815100117.
Shmelev, A. N., E-mail: shmelan@mail.ru, Kulikov, G. G., E-mail: ggkulikov@mephi.ru, Kurnaev, V. A., E-mail: kurnaev@yandex.ru, Salahutdinov, G. H., E-mail: saip07@mail.ru, Kulikov, E. G., E-mail: egkulikov@mephi.ru, & Apse, V. A., E-mail: apseva@mail.ru. Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors. United States. doi:10.1134/S1063778815100117.
Shmelev, A. N., E-mail: shmelan@mail.ru, Kulikov, G. G., E-mail: ggkulikov@mephi.ru, Kurnaev, V. A., E-mail: kurnaev@yandex.ru, Salahutdinov, G. H., E-mail: saip07@mail.ru, Kulikov, E. G., E-mail: egkulikov@mephi.ru, and Apse, V. A., E-mail: apseva@mail.ru. 2015. "Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors". United States. doi:10.1134/S1063778815100117.
@article{osti_22471976,
title = {Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors},
author = {Shmelev, A. N., E-mail: shmelan@mail.ru and Kulikov, G. G., E-mail: ggkulikov@mephi.ru and Kurnaev, V. A., E-mail: kurnaev@yandex.ru and Salahutdinov, G. H., E-mail: saip07@mail.ru and Kulikov, E. G., E-mail: egkulikov@mephi.ru and Apse, V. A., E-mail: apseva@mail.ru},
abstractNote = {Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the {sup 231}Pa–{sup 232}U–{sup 233}U–Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of {sup 232}U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.},
doi = {10.1134/S1063778815100117},
journal = {Physics of Atomic Nuclei},
number = 10,
volume = 78,
place = {United States},
year = 2015,
month =
}
  • The possible role of available thorium resources of the Russian Federation in utilization of thorium in the closed (U–Pu)-fuel cycle of nuclear power is considered. The efficiency of application of fusion neutron sources with thorium blanket for economical use of available thorium resources is demonstrated. The objective of this study is the search for a solution of such major tasks of nuclear power as reduction of the amount of front-end operations in the nuclear fuel cycle and enhancement of its protection against uncontrolled proliferation of fissile materials with the smallest possible alterations in the fuel cycle. The earlier results aremore » analyzed, new information on the amount of thorium resources of the Russian Federation is used, and additional estimates are made. The following basic results obtained on the basis of the assumption of involving fusion reactors with Th-blanket in future nuclear power for generation of the light uranium fraction {sup 232+233+234}U and {sup 231}Pa are formulated. (1) The fuel cycle would shift from fissile {sup 235}U to {sup 233}U, which is more attractive for thermal power reactors. (2) The light uranium fraction is the most “protected” in the uranium fuel component, and being mixed with regenerated uranium, it would become reduced-enrichment uranium fuel, which would relieve the problem of nonproliferation of the fissile material. (3) The addition of {sup 231}Pa into the fuel would stabilize its neutron-multiplying properties, thus making it possible to implement a long fuel residence time and, as a consequence, increase the export potential of the whole nuclear power technology. (4) The available thorium resource in the vicinity of Krasnoufimsk is sufficient for operation of the large-scale nuclear power industry of the Russian Federation with an electric power of 70 GW for more than one quarter of a century. The general conclusion is that involvement of a small number of fusion reactors with Th-blanket in the future nuclear power industry of the Russian Federation would to a large extent solve its problems and increase its export potential.« less
  • The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel canmore » be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.« less
  • Plasma size and other basic performance parameters for 1000-MW(electric) power production are calculated with the blanket energy multiplication factor, the M value, as a parameter. The calculational model is base don the International Thermonuclear Experimental Reactor (ITER) physics design guidelines and includes overall plant power flow. Plasma size decreases as the M value increases. However, the improvement in the plasma compactness and other basic performance parameters, such as the total plant power efficiency, becomes saturated above the M = 5 to 7 range. THus, a value in the M = 5 to 7 range is a reasonable choice for 1000-MW(electric)more » hybrids. Typical plasma parameters for 1000-MW(electric) hybrids with a value of M = 7 are a major radius of R = 5.2 m, minor radius of a = 1.7 m, plasma current of I{sub p} = 15 MA, and toroidal field on the axis of B{sub o} = 5 T. The concept of a thermal fission blanket that uses light water as a coolant is selected as an attractive candidate for electricity-producing hybrids. An optimization study is carried out for this blanket concept. The result shows that a compact, simple structure with a uniform fuel composition for the fissile region is sufficient to obtain optimal conditions for suppressing the thermal power increase caused by fuel burnup. The maximum increase in the thermal power is +3.2%. The M value estimated from the neutronics calculations is {approximately}7.0, which is confirmed to be compatible with the plasma requirement. These studies show that it is possible to use a tokamak fusion core with design requirements similar to those of ITER for a 1000-MW(electric) power reactor that uses existing thermal reactor technology for the blanket. 30 refs., 22 figs., 4 tabs.« less
  • The hybrid fusion reactor is becoming an interesting and promising model. In the present Note, a method for controlling the breeding-fission ratio is investigated. Since /sup 238/U fission occurs mainly with fast neutrons and breeding occurs with intermediate and slow neutrons, an optimal ratio can be obtained by partial slowing down of the original 14.9-MeV neutrons. This is done using iron as the moderator. Uranium samples were irradiated with 14.9-MeV neutrons from a deuterium-tritium reaction with iron layers of various thicknesses betweeen the samples and the neutron source. It was found that with a relatively thin layer of iron (12more » cm), any breeding-fission ratio can be obtained within a range of two decades. The breeding rate changes by only 50% when the iron-slab thickness changes from 0 to 12 cm, while the fission rate follows (more or less) the 14-MeV neutron flux and drops by more than two decades. Good agreement was obtained between the measurement and the calculated results.« less
  • This study focused on creating a new tristructural isotropic (TRISO) coated particle fuel performance model and demonstrating the integration of this model into an existing system of neutronics and heat transfer codes, creating a user-friendly option for including fuel performance analysis within system design optimization and system-level trade-off studies. The end product enables both a deeper understanding and better overall system performance of nuclear energy systems limited or greatly impacted by TRISO fuel performance. A thorium-fueled hybrid fusion-fission Laser Inertial Fusion Energy (LIFE) blanket design was used for illustrating the application of this new capability and demonstrated both the importancemore » of integrating fuel performance calculations into mainstream design studies and the impact that this new integrated analysis had on system-level design decisions. A new TRISO fuel performance model named TRIUNE was developed and verified and validated during this work with a novel methodology established for simulating the actual lifetime of a TRISO particle during repeated passes through a pebble bed. In addition, integrated self-consistent calculations were performed for neutronics depletion analysis, heat transfer calculations, and then fuel performance modeling for a full parametric study that encompassed over 80 different design options that went through all three phases of analysis. Lastly, side studies were performed that included a comparison of thorium and depleted uranium (DU) LIFE blankets as well as some uncertainty quantification work to help guide future experimental work by assessing what material properties in TRISO fuel performance modeling are most in need of improvement. A recommended thorium-fueled hybrid LIFE engine design was identified with an initial fuel load of 20MT of thorium, 15% TRISO packing within the graphite fuel pebbles, and a 20cm neutron multiplier layer with beryllium pebbles in flibe molten salt coolant. It operated at a system power level of 2000 MW th, took about 3.5 years to reach full plateau power, and was capable of an End of Plateau burnup of 38.7 %FIMA if considering just the neutronic constraints in the system design; however, fuel performance constraints led to a maximum credible burnup of 12.1 %FIMA due to a combination of internal gas pressure and irradiation effects on the TRISO materials (especially PyC) leading to SiC pressure vessel failures. The optimal neutron spectrum for the thorium-fueled blanket options evaluated seemed to favor a hard spectrum (low but non-zero neutron multiplier thicknesses and high TRISO packing fractions) in terms of neutronic performance but the fuel performance constraints demonstrated that a significantly softer spectrum would be needed to decrease the rate of accumulation of fast neutron fluence in order to improve the maximum credible burnup the system could achieve.« less