Fission-suppressed blankets for fissile fuel breeding fusion reactors

Two blanket concepts for deuterium-tritium (DT) fusion reactors are presented which maximize fissile fuel production while at the same time suppress fission reactions. By suppressing fission reactions, the reactor will be less hazardous, and therefore easier to design, develop, and license. A fusion breeder operating a given nuclear power level can produce much more fissile fuel by suppressing fission reactions. The two blankets described use beryllium for neutron multiplication. One blanket uses two separate circulating molten salts: one salt for tritium breeding and the other salt for U-233 breeding. The other uses separate solid forms of lithium and thorium for breeding and helium for cooling.Nuclear power is the sum of fusion (D + T 14 MeV neutron+ 3.5 MeV alpha) power plus additional power from neutron-induced reactions in the blanket.     

利用D-T聚变中子源转化长寿命裂变产物的中子学可行性研究

从中子学角度研究长寿命裂变产物在Tokamak型D-T聚变堆包层中转化的可行性.提出了用可裂变Pu增殖中子的混合包层转化方案,研制了相应的燃耗计算程序及数据库,并对所提方案进行了计算和分析.结果表明,在可预见的聚变堆芯技术条件下,所研究的概念性包层可对长寿命裂变产物进行有效转化.     

Status of gas-cooled fast breeder reactor programs

Most gas-cooled fast breeder reactor (GCFR) programs in Europe and the US are now coordinated and focused on a 300 MW(e) GCFR demonstration plant program. Except for venting and artificial surface roughening, GCFR fuel is similar to liquid metal fast breeder reactor (LMFBR) fuel and operates under nearly identical conditions. The primary helium system is integrated within a PCRV like all large gas-cooled thermal reactors, with three main loops and three auxiliary loops. Design and safety studies and various experiments, including heat transfer, irradiation, and critical experiments, indicate that most feasibility questions have been answered and a demonstration plant could be in operation within 12 years. This could be followed in the mid-1990s by a large-size GCFR with a doubling time of about 10 years fueled by (UO₂---PuO₂) and producing either ²³³U in thorium blankets as fuel for advanced converters or plutonium in depleted uranium blankets.     

Neutronics study for the installation of the LBVM in the medium flux area of IFMIF

IFMIF (International Fusion Materials Irradiation Facility) will be a fusion dedicated facility producing a large amount of neutrons with the appropriate energy spectrum to test materials and subcomponents for DEMO and future Fusion Power Plants.While the high flux area of IFMIF will be devoted to reduced activation structural materials for first wall and blanket, the medium flux area will be dedicated to functional materials for breeder blankets. In particular, the Liquid Breeder Validation Module (LBVM), will host experiments related with functional materials for liquid breeder blankets. Since IFMIF neutron spectra have been intended to fit the most irradiated areas of a fusion reactor in the high flux area, the irradiation conditions in the LBVM placed in the medium flux area of IFMIF have been assessed. The effect of some neutron shifter/reflector components to optimize the neutron spectra have been evaluated in order to find out the proper irradiation conditions for functional materials for liquid breeder blankets.Therefore, the objective of this report is to summarize the neutronic calculations developed to evaluate the viability of IFMIF neutron source to perform relevant irradiation experiments on functional materials for liquid breeder blanket concept for future nuclear fusion power reactors (ITER, DEMO). The irradiation parameters evaluated for this purpose are: the tritium production for liquid breeder material (Pb–17Li) and the damage dose (dpa) and gas production to damage dose ratios for Al₂O₃ and SiC functional materials.The main conclusion is that, it is possible to perform relevant irradiation experiments on functional materials for liquid breeder blanket concept for the future nuclear fusion reactor DEMO. Nevertless, the use of some shifter components will be needed to optimize some irradiation parameters.     

Neutronic analysis of a thorium fusion breeder with enhanced protection against nuclear weapon proliferation

The fissile breeding capability of a (D,T) fusion-fission (hybrid) reactor fueled with thorium is analyzed to provide nuclear fuel for light water reactors (LWRs). Three different fertile material compositions are investigated for fissile fuel breeding: (1) ThO₂; (2) ThO₂ denaturated with 10% natural-UO₂ and (3) ThO₂ denaturated with 10% LWR spent fuel. Two different coolants (pressurized helium and Flibe ‘Li₂BeF₄’) are selected for the nuclear heat transfer out of the fissile fuel breeding zone. Depending on the type of the coolant in the fission zone, fusion power plant operation periods between 30 and 48 months are evaluated to achieve a fissile fuel enrichment quality between 3 and 4%, under a first-wall fusion neutron energy load of 5 MW/m² and a plant factor of 75%. Flibe coolant is superior to helium with regard to fissile fuel breeding. During a plant operation over four years, enrichment grades between 3.0 and 5.8% are calculated for different fertile fuel and coolant compositions. Fusion breeder with ThO₂ produces weapon grade ²³³U. The denaturation of the ²³³U fuel is realized with a homogenous mixture of 90% ThO₂ with 10% natural-UO₂ as well as with 10% LWR spent nuclear fuel. The homogenous mixture of 90% ThO₂ with 10% natural-UO₂ can successfully denaturate ²³³U with ²³⁸U. The uranium component of the mixture remains denaturated over the entire plant operation period of 48 months. However, at the early stages of plant operation, the generated plutonium component is of weapon grade quality. The plutonium component can be denaturated after a plant operation period of 24 and 30 months in Flibe cooled and helium cooled blankets, respectively. On the other hand, the homogenous mixture of 90% ThO₂ with 10% LWR spent nuclear fuel remains non-prolific over the entire period for both, uranium and plutonium components. This is an important factor with regard to international safeguarding.     

Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80 tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000 MWth output and tritium self-sufficiency (TBR > 1.15), while the required maximum fusion powers are lower than 500 MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%.     

Power Flattening in ARIES-RS Fusion Breeder Reactor Using Mixed Fuels

This study presents the possibility of the power flattening in the ARIES-RS breeder reactor using mixed (Th,U)C or (Th,U)N fuels. Two different types of mixing, namely, homogeneous mixing (HM) and linear mixing (LM) were used to investigate the uniformity of fission power distribution through the fuel zone. In HM, fraction of uranium content were kept constant in all rows of the fuel zone whereas, in LM the fraction of the uranium were linearly increased from the first to last fuel row in the fuel zone. Neutron transport calculations were performed with the aid of the SCALE4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S₈–P₃ approximation. Flat fission power distribution was maintained successfully for the blanket using linearly mixed fuels. However, the fission density profile was not uniform in the blanket with homogeneously mixed fuels. It decreased exponentially form the 1st to 10th fuel row.     

Neutronic analysis of PROMETHEUS reactor fueled with various compounds of thorium and uranium

In this study, neutronic performance of the DT driven blanket in the PROMETHEUS-H (heavy ion) fueled with different fuels, namely, ThO₂, ThC, UO₂, UC, U₃Si₂ and UN is investigated. Helium is used as coolant, and SiC is used as cladding material to prevent fission products from contaminating coolant and direct contact fuel with coolant in the blanket. Calculations of neutronic data per DT fusion neutron are performed by using SCALE 4.3 Code. M (energy multiplication factor) changes from 1.480 to 2.097 depending on the fuel types in the blanket under resonance-effect. M reaches the highest value in the blanket fueled with UN. Therefore, the investigated reactor can produce substantial electricity in situ. UN has the highest value of ²³⁹Pu breeding capability among the uranium fuels whereas UO₂ has the lowest one. ²³⁹Pu production ratio changes from 0.119 to 0.169 according to the uranium fuel types, and ²³³U production values are 0.125 and 0.140 in the blanket fueled with ThO₂ and ThC under resonance-effect, respectively. Heat production per MW (D,T) fusion neutron load varies from 1.30 to 7.89 W/cm³ in the first row of fissile fuel breeding zone depending on the fuel types. Heat production attains the maximum value in the blanket fueled with UN. Values of TBR (tritium breeding ratio) being one of the most important parameters in a fusion reactor are greater than 1.05 for all type of fuels so that tritium self-sufficiency is maintained for DT fusion driver. Values of peak-to-average fission power density ratio, Γ, are in the range of 1.390 and ∼1.476 depending on the fuel types in the blanket. Values of neutron leakage out of the blanket for all fuels are quite low due to SiC reflector. The maximum neutron leakage is only ∼0.025. Consequently, for all cases, the investigated reactor has high neutronic performance and can produce substantial electricity in situ, fissile fuel and tritium required for (D,T) fusion reaction.     

Neutronic performance of new coolants in a fusion–fission (hybrid) reactor

Selection of coolant used in the fuel zone of a fusion–fission (hybrid) reactor affects the neutronic performance of the blanket much. Recently, two coolants namely, Flinabe and Li₂₀Sn₈₀ have been investigated to use in fusion reactors as tritium breeder and energy carrier due to their advantages of low melting point, low vapor pressure. In this study, neutronic performance of these coolants in a hybrid reactor using Canada Deuterium Uranium Reactor (CANDU) spent fuel was investigated for an operation period of 48 months. And also that of natural lithium and Flibe was also examined for comparison. Neutron transport calculations were conducted on a simple experimental hybrid blanket in a cylindrical geometry with the help of the SCALE4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S₈–P₃ approximation.     

聚变—裂变混合堆及其在我国核能发展中的作用

本文概要介绍聚变和聚变-裂变混合堆基本原理及其作用。聚变-裂变混合堆可以为压水堆或快堆提供充足的核燃料。它和压水堆或快堆组成的系统具有经济可行性。在解决我国核能发展中燃料短缺问题和促进纯聚变能源的发展方面可望发挥重要的作用。     

The Spallator and APEX Nuclear Fuel Cycle a New Option for Nuclear Power

A new nuclear fuel cycle is described which provides a long term supply of nuclear fuel for the thermal LWR nuclear power reactors and eliminates the need for long-term storage of radioactive waste. Fissile fuel is produced by the Spallator which depends on the production of spallation neutrons by the interaction of high energy (1 to 2 GeV) protons on a heavy metal target. The neutrons are absorbed in a surrounding natural uranium or thorium blanket in which fissile Pu-239 or U-233 is produced. Advances in linear accelerator technology makes it possible to design and construct a high beam current continuous wave proton linac for production purposes. The target is similar to a sub-critical reactor and produces heat which is converted to electricity for supplying the linac. The Spallator is a self-sufficient fuel producer, which can compete with the fast breeder. The APEX fuel cycle depends on recycling the transuranics and long-lived fission products while extracting the stable and short-lived fission products when reprocessing the fuel. Transmutation and decay within the fuel cycle and decay of the short-lived fission products external to the fuel cycle eliminates the need for long-term geological age storage of fission product waste.     

Design optimization of first wall and breeder unit module size for the Indian HCCB blanket module

The Indian test blanket module(TBM) program in ITER is one of the major steps in the Indian fusion reactor program for carrying out the RD activities in the critical areas like design of tritium breeding blankets relevant to future Indian fusion devices(ITER relevant and DEMO).The Indian Lead–Lithium Cooled Ceramic Breeder(LLCB) blanket concept is one of the Indian DEMO relevant TBM,to be tested in ITER as a part of the TBM program.Helium-Cooled Ceramic Breeder(HCCB) is an alternative blanket concept that consists of lithium titanate(Li_2TiO_3) as ceramic breeder(CB) material in the form of packed pebble beds and beryllium as the neutron multiplier.Specifically,attentions are given to the optimization of first wall coolant channel design and size of breeder unit module considering coolant pressure and thermal loads for the proposed Indian HCCB blanket based on ITER relevant TBM and loading conditions.These analyses will help proceeding further in designing blankets for loads relevant to the future fusion device.     

聚变-裂变混合堆水冷包层中子物理性能研究

研究直接应用国际热核聚变实验堆（ITER）规模的聚变堆作为中子驱动源,采用天然铀为初装核燃料,并采用现有压水堆核电厂成熟的轻水慢化和冷却技术,设计聚变-裂变混合堆裂变及产氚包层的技术可行性。应用MCNP与Origen2相耦合的程序进行计算分析,研究不同核燃料对包层有效增殖系数、氚增殖比、能量放大系数和外中子源效率等中子物理性能的影响。计算分析结果显示,现有核电厂广泛使用的UO₂核燃料以及下一代裂变堆推荐采用的UC、UN和U₉₀Zr₁₀等高性能陶瓷及合金核燃料作为水冷包层的核燃料,都能满足以产能发电为设计目标的新型聚变 裂变混合堆能量放大倍数的设计要求,但只有UC和U₉₀Zr₁₀燃料同时满足聚变燃料氚的生产与消耗自持的要求。研究结果对进一步研发满足未来核能可持续发展的新型聚变-裂变混合堆技术具有潜在参考价值。     

Study of thorium–uranium based molten salt blanket in a fusion–fission hybrid reactor

Not only solid fuels, but also liquid fuels can be used for the fusion–fission symbiotic reactor blanket. The operational record of the molten salt reactor with F–Li–Be was very successful, so the F–Li–Be blanket was chosen for research. The molten salt has several features which are suited for the fusion–fission applications.The fuel material uranium and thorium were dissolved in the F–Li–Be molten salt. A combined program, COUPLE, was used for neutronics analysis of the molten salt blanket. Several cases have been calculated and compared. Not only the influence of the different fuels have been studied, but also the thickness of the molten salt, and the concentration of the ⁶Li in the molten salt.Preliminary studies indicate that when thorium–uranium–plutonium fuels were added into a F–Li–Be molten salt blanket and with a component of 71% LiF–2% BeF₂–13.5% ThF₄–8.5% UF₄–5% PuF₃, and also with the molten salt thickness of 40 cm and natural concentration of ⁶Li, the appropriate blanket energy multiplication factor and TBR can be obtained.The result shows that thorium–uranium molten salt can be used in the blanket of a fusion–fission symbiotic reactor. The research on the molten salt blanket must be valuable for the design of fusion–fission symbiotic reactor.     

Energy sustainability and economic stability with Breed and Burn reactors

The purpose of the present study is to evaluate the impact a successful development of Breed and Burn (B&B) fast reactors and their fuel reconditioning technologies could have on the uranium ore utilization, uranium enrichment capacity, nuclear waste and energy security. It is found that a successful development of B&B reactors will offer 40-folds increase in the uranium ore utilization versus that presently achieved. A successful development of a fuel reconditioning technology could increase the attainable uranium utilization to 100-folds its present value. The growth rate of the installed capacity of B&B reactors possible to achieve using the “spawning” mode of operation is estimated to be nearly 4% per year. The amount of natural uranium required for starting a fleet of B&B reactors that will reach an electricity generation capacity of 1000 GWe by the end of this century is estimated to be the equivalent of 10 years of supply to the presently operating commercial fleet of LWRs in the US (86 GWe). No natural uranium and no enrichment capacity will be required to support this fleet beyond the later part of this century. The energy value of the depleted uranium stockpiles (“waste”) that will be accumulated in the US by that time is equivalent to, when used in the B&B reactors, up to 20 centuries of the total 2010 supply of electricity in the USA. It is therefore concluded that a successful development of B&B reactors and associated fuel reconditioning could provide a great measure of energy security, proliferation resistance and cost stability.     

Neutronic analysis of an inertial fusion energy breeder with SiCf/SiC

Neutronic performance of a blanket driven ICF (Inertial confinement fusion) neutron based on SiC_f/SiC composite material is investigated for fissile fuel breeding. The investigated blanket is fueled with ThO₂ and cooled with natural lithium or (LiF)₂BeF₂ or Li₁₇Pb₈₃ or ⁴He coolant. MCNP4B Code is used for calculations of neutronic data per DT neutron. Calculations have show that values of TBR (tritium breeding ratio) being one of the main neutronic paremeters of fusion reactors are greater than 1.05 in all type of coolant, and the breeder hybrid reactor is self-sufficient in the tritium required for the DT fusion driver. Calculations show that natural lithium coolant blanket has the highest TBR (1.298) and M (fusion energy multiplication) (2.235), Li₁₇Pb₈₃ coolant blanket has the highest FFBR (fissile fuel breeding ratio) (0.3489) and NNM (net neutron multiplication) (1.6337). ⁴He coolant blanket has also the best Γ (peek-to-average fission power density ratio) (1.711). Values of neutron leakage out of the blanket in all type of coolants are quite low due to SiC reflector and B₄C shielding.     

MCORGS及次临界能源堆包层概念研究进展

次临界能源堆是以能源供应为目的的一种聚变裂变混合堆,以聚变驱动,天然铀为裂变燃料,轻水为冷却剂。本文针对该混合堆开发了基于MCNP与ORIGENS的三维中子输运燃耗耦合程序MCORGS,分析了包层三维中子学模型。提出简化干法后处理,设想利用衰变热将乏燃料加热到2 100K,将沸点低于该温度的裂变产物挥发去除。计算了包层各区材料每年发生的原子移位数,建议采用10a左右的换料周期,乏燃料经后处理后可多次复用。第1个寿期内氚增殖比TBR平均约1.15,包层能量放大倍数M平均约12;第2~9个寿期内TBR平均约1.35,M平均约18。利用流体动力学程序完成了包层CAD模型建立、网格划分及稳态传热计算分析,各区材料的最高温度均低于许用温度并有较大裕量。     

The closed on-site fuel cycle of the brest reactors

The BREST fast reactor with nitride fuel and lead coolant is being developed as a reactor of new generation, which has to meet a set of requirements placed upon innovative reactors, namely efficient use of fuel resources, nuclear, radiation and environmental safety, proliferation resistance, radwaste treatment and economic efficiency. Mixed uranium-plutonium mononitride fuel composition allows supporting in BREST reactor CBR≈1. It is not required to separate plutonium to produce “fresh” fuel. Coarse recovered fuel purification of fission products is allowed (residual content of FPs may be in the range of 10⁻² – 10⁻³ of their content in the irradiated fuel). High activity of the regenerated fuel caused by minor actinides is a radiation barrier against fuel thefts. The fuel cycle of the BREST-type reactors “burns” uranium-238, which must be added to the fuel during reprocessing. Plutonium is not extracted during reprocessing being a part of fuel composition, thus exhibiting an important nonproliferation feature.

The radiation equivalence between natural uranium consumed by the BREST NPP closed system and long-lived high-level radwaste is provided by actinides (U, Pu, Am) transmutation in the fuel and long-lived products (I, Tc) transmutation in the blanket. The high-level waste must be stored for approximately 200 years to reduce its activity by the factor of about 1000.

The design of the building and the entire set of the fuel cycle equipment has been completed for the demonstration BREST-OD-300 reactor, which includes all main features of the BREST-type reactor on-site closed fuel cycle.     

Optimum utilization of nuclear fuel with gas and vapor core reactors

Gas and Vapor Core Reactors (G/VCR) are externally reflected and moderated nuclear energy systems fueled by stable uranium compound in gaseous or vapor phase. In G/VCR systems the functions of fuel and coolant are combined and the reactor outlet temperature is not constrained by solid fuel-cladding temperature limitations. G/VCRs can potentially provide the highest reactor and cycle temperature among all existing or proposed fission reactor designs. Furthermore, G/VCR systems feature a low inventory and fully integrated fuel cycle with exceptional sustainability and safety characteristics. With respect to fuel utilization, there is practically no fuel burn-up limit for gas core reactors due to continuous recycling of the fuel. Owing to flexibility in nuclear design characteristics of cavity reactors, a wide range of conversion ratio from almost solely a burner to a breeder is achievable. The continuous recycling of fuel in G/VCR systems allows for continuous burning and transmutation of actinides without removing and reprocessing of the fuel. The only waste product at the backend of the gas core reactors' fuel cycle is fission fragments that are continuously separated from the fuel. As a result the G/VCR systems do not require spent fuel storage or reprocessing.

G/VCR systems also feature outstanding proliferation resistance characteristics and minimum vulnerability to external threats. Even for comparable spectral characteristic, gas core reactors produce fissile plutonium two orders of magnitude less than Light Water Reactors (LWRs). In addition, the continuous transmutation and burning of actinides further reduces the quality of the fissile plutonium inventory. The low fuel inventory (about two orders of magnitude lower than LWRs for the same power generation level) combined with continuous burning of actinides, significantly reduces the need for emergency planning and the vulnerability to external threats. Low fuel inventory, low fuel heat content, and online separation of fission fragments are among the key constituent safety features of G/VCR systems.     

Preliminary design of hybrid energy reactor and integral neutron experiments

We propose a preliminary design for a fusion–fission hybrid energy reactor (FFHER), based on current fusion science and technology (with some extrapolations forward from ITER) and well-developed fission technology. We list design rules and put forward a primary concept blanket, with uranium alloy as fuel and water as coolant. The uranium fuel can be natural uranium, LWR spent fuel, or depleted uranium. The FFHER design can increase the utilization rate of uranium in a comparatively simple way to sustain the development of nuclear energy. We study the interaction between the fusion neutron and the uranium fuel with the aim of to achieving greater energy multiplication and tritium sustainability. We also review other concept hybrid reactor designs. We design integral neutron experiments in order to verify the credibility of our proposed physical design. The combination of this program of research with the related thermal hydraulic design, alloy fuel manufacture, and nuclear fuel cycle programs provides the science and technology foundation for the future development of the FFHER concept in China.