Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation’s energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.     

The sensitivity of fuel cycle performance to separation efficiency

Reprocessing separation efficiency is a major design variable in the implementation of advanced fuel cycles as it affects waste disposal requirements, fuel fabrication, system economics, and other fuel cycle system characteristics. Using a newly developed, physics-based integrated fuel cycle systems analysis model, this study investigated the impact of varying reprocessing separation efficiencies on fuel cycle cost (FCC), proliferation resistance and repository impact. Repository impact was captured by the disposal facility capacity governed by thermal output, the projected dose rate, mass inventory, and waste toxicity index. The coupled systems analysis model included fast reactor simulation tool to analyze the depletion in the fast reactor and the requirements for the fresh fuel in transient and equilibrium states. In this calculation, the feedback between separation efficiencies and fresh and discharged fuel compositions was dynamically accounted for. The new systems model was benchmarked against published results and used to investigate a single-tier nuclear fuel cycle scenario in which light water reactors (LWRs) and 0.5 transuranic (TRU) conversion ratio (CR) sodium-cooled fast reactors are deployed in an equilibrium that results in zero net TRU production. The results indicated that fuel cycle system performance is significantly affected by the changes in partitioning strategies and elemental separation efficiency in reprocessing plants. Moreover, the effect of varying separation efficiencies on reactor performance, fuel cycle mass balances and economic performance are discussed.     

Nuclear mechanism for TRU burning oxide fueled core in Advanced Recycling Reactor

This paper presents an approximation approach to predict the core characteristics based on parametric survey and an analysis of nuclear mechanism in a conceptual nuclear design for enhanced transuranics (TRU) burning mixed oxide fueled and sodium cooled fast reactor which can be realized in the near future. The design study of Advanced Recycling Reactor was conducted in the context of the program for the industry in Global Nuclear Energy Partnership initiatives, including a core in the first plant for demonstration and cores with enhanced TRU burning capability for the future plants. Both concepts for the first plant; low core height and large volume fraction of structure are deployed, seeking small TRU conversion ratio (CR)^* and small void reactivity which are crucial in the design, but different approaches. In this paper, the TRU CR and the sodium void reactivity have been approximated with a single equation in these concepts, based on the theoretical formula related to the chain reaction in the reactor and the calculation results. Shortening the core height and increasing the structure volume fraction will enhance TRU enrichment through increased neutron leakage and capture, which will reduce a ratio of U-238 to sum of Pu-239 and Pu-241 so that TRU CR decreases from 0.78 to 0.57. A small ratio of sodium loss to plutonium fissile will be effective also in the reduction of positive reactivity effect by spectral hardening. On the other hand, when this ratio and geometrical buckling of flux reduce, negative contribution by the neutron leakage becomes small. Theses relations related to the positive void reactivity have been formularized by the approximation with few parameters within several percent respectively as well as the TRU CR, indicating that one of dominating parameters is the ratio of sodium loss to plutonium fissile in the void reactivity at large fast reactors. * = (1 − net loss of TRU/loss of heavy metal).     

Conceptual core design of Advanced Recycling Reactor based on mature technologies

This paper presents about comprehensive investigations into Advanced Recycling Reactor (ARR) based on existing and/or mature technologies (called “Early ARR”), aiming transuranics (TRU) burning and considering harmonization of TRU burning capability, technology readiness, economy and safety. The ARR is a 500 MW_e (1180 MW_th) oxide fueled sodium cooled fast reactor, which the low core height of 70 cm and the large structure volume fraction with 1.0 mm of cladding thickness to tube wall have been chosen among 14 candidate concepts to reduce the TRU conversion ratio (CR) and the void reactivity, taking technology readiness into account. As a result of nuclear calculation, the ARR has TRU burning capability from 19 to 21 kg/TW_thh and is sustainable in recycling. And the ARR can accept several kinds of TRU; the LWR uranium oxide fuels, LWR-MOX used nuclear fuel, and TRU recycled in this fuel cycle and the ARR is also flexible in TRU management in ways that it can transform from TRU CR of 0.56 to breeding ratio (BR) of 1.03. In addition, it has been confirmed that the ARR core conforms to the set design requirements; the void reactivity, the maximum linear heat rate, and the shutdown margin of reactivity control system. It has been confirmed that the closed fuel cycle with the ARR plants of 180 GW_th will not release TRU outside and generate more electricity by 65% compared with the present nuclear power system in the US, curbing the risk of nuclear proliferation. Thus the study can conclude that the Early ARR is able to close nuclear fuel cycle, using mature technologies and has features of the sustainability in recycling, and the accommodation of almost all the TRU at present and in the future, and the flexibility in TRU management with breakeven core.     

Physics features comparison of TRU burners: Fusion/Fission Hybrids,Accelerator-Driven Systems and low conversion ratio critical fast reactors

This paper compares different types of TRU burners, sub-critical (as Accelerator-Driven Systems and Fusion Fission Hybrids) but also critical, low conversion ratio, fast reactors. To make a significant comparison, it is specified for which objective and within which strategy these systems can be envisaged. Beside intrinsic cost parameters, the associated fuel cycle issues can prove to be crucial for their deployment.     

A feasibility study on FP incineration for self-consistent nuclear energy system (SCNES)

A fast reactor core and fuel cycle concept is discussed for the future “Self-Consistent Nuclear Energy System (SCNES)” concept. The present study mainly discussed long-lived fission products (LLFPs) burning capability and recycle scheme in the framework of metal fuel fast reactor cycle, aiming at the goals for fuel breeding capability and confinement for TRU and radio-active FPs within the system. Combining neutron spectrum-shift for target sub-assemblies and isotope separation using tunable laser, LLFP burning capability is enhanced. This result indicates that major LLFPs can be treated in the additional recycle schemes to avoid LLFP accumulation along with energy production. In total, the proposed fuel cycle is a candidate for realizing SCNES concept.     

DRAGON&DONJON程序在MSR中堆芯燃耗计算的适用性

DRAGONDONJON组件-堆芯"两步法"程序通过合理简化,理论可适用于任何堆芯与工况。使用蒙特卡罗方法 RMC(Reactor Monte Carlo code)、MCNP(Monte Carlo Neutron Particle transport code)程序验证DRADON程序是否能够承担快/热谱型熔盐堆(Molten Salt Reactor,MSR)焚烧TRU、Th U燃料燃耗计算。选出熔盐增殖堆(Molten Salt Breeder Reactor,MSBR)与熔盐锕系元素再循环和嬗变堆(Molten Salt Advanced Reactor Transmuter,MOSART)堆型进行比较,同时分别利用RMC程序验证DRAGON程序组件燃耗计算的准确性,利用MCNP程序验证DRAGON程序组件均匀化方法以及DONJON程序截面调用和程序全堆扩散的准确性。结果表明,组件燃耗计算中,TRU和Th U燃料满足燃耗计算要求;堆芯临界计算中,快/热谱堆芯计算误差均小于0.001。证明DRADON程序可以胜任快、热谱型MSR焚烧TRU、Th U燃料的物理计算任务。     

加速器驱动系统的燃料选择

对装载不同增殖材料的现实加速器驱动系统(ADS)的安全及嬗变超铀核素特性进行研究。分别 以(U,TRU)O2和(Th,TRU)O2作为堆芯燃料,先用LAHET和MCNP程序对ADS进行稳态模拟计 算,再耦合MCNP和ORIGEN2程序计算燃耗过程中的核素密度变化。结果显示,装载钍基燃料的 ADS对超铀核素的嬗变效果较好,且在燃耗过程中其反应性和质子流强波动较小;装载铀基燃料的 ADS则具有更安全的多普勒效应和缓发中子有效份额。总体来看,如果需要堆长时间安全嬗变超铀核 素,装载钍基燃料会取得更好的效果。     

Plutonium and Minor Actinides incineration options using innovative Na-cooled fast reactors: Impacting on phasing-out and on-going fuel cycles

The flexibility of innovative Na-cooled fast reactors for burning Pu and/or Minor Actinides (MA) is investigated with respect to different fuel cycle strategies. Under phasing-out conditions, the burner systems are used for reducing to a minimum level the accumulated TRansUranic (TRU) inventory, whereas when continuous use of nuclear energy is envisaged (on-going case), burner systems may be dedicated to MA management only.As an example of a phasing-out case, the accumulated German TRU inventory (at 2022) is assumed to be transmuted in a chosen time period of 150 years. For this purpose, two different burner fast reactors concepts, developed at KIT, are deployed in a Partitioning and Transmutation based fuel cycle. The effects are analyzed in order to confirm the behavior expected by the neutronics studies and to provide a basis for further optimization of the scenarios with respect to a number of reactors, deployment paces and fuel compositions.Additionally the performance of the MA burner is assessed to provide an effective MA mass stabilization in case of a continuous use of nuclear energy. Preliminary results are compared with those of past studies based on the European Sodium-cooled Fast Reactor.     

Analysis of Cell Configuration Effects on Nuclear Characteristics in a Lead-Bismuth Cooled Core for TRU Incineration

Sensitivity of the core characteristics to the fuel pin cell parameters change is analyzed for a lead-bismuth cooled reactor to incinerate transuranic nuclides. The pitch-to-diameter ratio is changed for a parametric study to investigate the effects of the coolant-to-fuel ratio. Not only the Zr-based fuel of TRU+Zr but also the Th-based fuel of TRU+Th+Zr is considered in order to investigate the sensitivity of nuclear characteristics of the fuel pin cell to neutron energy spectrum as well as effects of the fuel type on the core performance. For the sensitivity analyses, the neutron spectrum, the criticality performance parameters, and the non-fissile actinides destruction factor are evaluated. The obtained results clarify the unique property of nuclear characteristics of the fuel pin cell and give some useful information for design optimization of a lead-bismuth cooled reactor for TRU transmutation.     

Update of the French R&D strategy on gas-cooled reactors

Gas-cooled reactors take up a strong second role in France's R&D strategy on future nuclear energy systems as priority was given in 2005 to fast neutron reactors with multiple-recycle for their potential to optimally use uranium resource and minimize the long term burden of radioactive waste. Owing to the European past experience on sodium-cooled fast reactors (SFRs), this reactor type was logically selected as reference for a new generation fast neutron reactor intended to be tested as a prototype in the 2020s and be ready for industrial deployment around 2040. At the same time, the potential merits of a gas fast reactor (GFR) with ceramic clad fuel for a safe management of cooling accident are acknowledged for the potential of this reactor type to resolve critical issues of liquid fast reactors (safety, operability and reparability). A pre-feasibility report on a first concept of GFR was issued in 2007 that summed-up results of a 5-year international R&D effort on GFR fuel technology, reactor design and operating transient analyses. This report established a global confidence in the feasibility of this concept and its potential for attractive performances. Furthermore, it suggested directions of R&D to generate by 2012 an updated concept with improved performances and taking better benefit from GFR specific technologies.A second activity on gas-cooled reactors originates from the current interest of CEA's industrial partner AREVA in high or very high temperature reactors (V/HTR) for supplying hydrogen, synthetic hydrocarbon fuels and process heat for the industry. This activity currently encompasses R&D on V/HTR key technologies such as particle fuel fabrication, high temperature compact heat exchangers and coupling technologies to various power conversion systems. R&D on V/HTR and GFR are synergistic in various respects. The GFR can be viewed as a more sustainable version of the VHTR and synergies exist in research on heat resisting materials, helium system technology and power conversion systems. Both reactors require active research in materials and spur developments of new metallic alloys and ceramics applicable to other advanced nuclear systems.     

Reactor physics ideas for large scale utilization of thorium in gas cooled reactors

The physics principles for maximizing the fertile to fissile conversion were used in developing reactor concepts for large scale utilization of thorium in thermal and fast reactors (Jagannathan & Pal, 2006; Jagannathan et al., 2008). It is recognized that these principles are very well suited for ‘He’ gas cooled reactors with graphite moderator since both helium gas coolant and the graphite moderator have low neutron absorption characteristics and thus gives better neutron economy. In this paper, these ideas are applied to the High Temperature Test Reactor (HTTR) core of Japan to assess its advantage over the present day gas cooled reactors. HTTR is helium cooled and graphite moderated system. Significant amount of thorium has been loaded in the HTTR core with some minimal changes in the existing core design. The modified design is called HTTR-M core.In the HTTR-M core, the fuel is changed from enriched UO₂ fuel to Pu in ThO₂ fuel. The locations of boron type burnable poison rods within each fuel assembly of HTTR are replaced by one cycle irradiated thoria rods. Also, the B₄C type control assembly around the HTTR core is replaced by fresh seedless thorium assembly. The fertile thoria assembly are scattered uniformly in the HTTR-M core. The equilibrium core of HTTR-M shows very small burnup reactivity swing. The core excess reactivity is ∼18 mk at BOC and reduces to 1 mk at 660 days. It is interesting to note that this small reactivity change is intrinsically achieved by the choice of seed and fertile dimensions and their contents without the use of burnable poison rods or mechanical control rods which are used in HTTR core. The burnup reactivity swing in the latter after using burnable poison is ∼100 mk. The fissile seed inventory ratio (FIR) in a fuel cycle is 0.90 as compared with 0.717 of HTTR core. Since ²³³U is a better fissile nuclide with highest ‘η’ value in thermal range, the above conversion ratio can be regarded as quite good.     

TRU燃料热谱MSR定期换料及嬗变特性数值研究

熔盐反应堆（MSR）燃料制备方便、中子经济性好、燃料管理灵活,具有直接利用轻水堆乏燃料中超铀核素（TRU）的潜力。本文通过优化燃料选取、栅格参数及燃料/石墨体积分数和去除裂变气体和惰性金属等方法,对TRU燃料热谱MSR堆芯寿期、TRU核素积存量、次锕系核素MA嬗变支持比和TRU焚毁率等进行计算分析,证明TRU燃料热谱MSR可实现长周期定期换料,减少在线换料的难度,同时对MA和TRU核素具有一定的嬗变能力,可降低乏燃料放射性毒性。      

Preliminary study on feasibility of large and small water cooled thorium breeder reactor in equilibrium states

Fuel breeding is one of the essential performances for a self-sustaining reactor system which can maintains the fuel sustainability while the reactor produces energy and consumes the fissile materials during operation. Thorium cycle shows some advantageous on higher breeding characteristics in thermal neutron spectrum region as shown in the Shippingport reactor and molten salt breeder reactor (MSBR) project. In the present study, the feasibility of large and small water cooled thorium breeder reactors is investigated under equilibrium conditions where the reactors are fueled by ²³³U–Th oxide and they adopts light water coolant as moderator. The key properties such as required enrichment, breeding capability, and initial fissile inventory are evaluated. The conversion ratio and fissile inventory ratio (FIR) are used for evaluating breeding performance. The results show the feasibility of breeding for small and large reactors. The breeding performance increases with increasing power output and lower power density. The small reactor may achieve the breeding condition when the fuel pellets' power density of about 22.5 W/cm³ and burnup of about 20 GWd/t.     

Hafnium clad fuels for fast spectrum BWRs

We show that by use of hafnium cladding, a fast neutron spectrum is achievable in the top of uprated BWRs. Monte Carlo calculations have been made for Hf clad inert matrix nitride and low fertile MOX fuels, with fuel segments located in the upper part of an uprated BWR, where the coolant void fraction exceeds 70%. The nitride fuel results in the hardest neutron spectrum, but the low fertile MOX fuel still yields fission probabilities for even neutron number nuclides similar to those of sodium cooled reactors. The inert matrix nitride fuel configuration yields high burning rates, permitting to stabilise TRU inventories with less than 50% BWR cores of the here suggested type in the power park. The core with low fertile MOX fuel is less efficient, but still a zero net producer of TRU. Fuel and coolant temperature feedbacks are affected by introduction of absorbing elements in the fuel, but remain within acceptable ranges for the low fertile MOX fuel. Although control rod worths are reduced, shutdown margins are sufficient to ensure sub-criticality in cold conditions. From a materials point of view, the behaviour of hafnium clad MOX fuel would be similar to zircalloy clad MOX fuel already used extensively in nuclear industry. Thus, if dynamic stability of the core can be ensured, the here proposed fuel may be considered as a low cost solution for transmutation of minor actinides on industrial scale.     

HFETR运行寿期和燃料元件装载量优化研究

文章建立了中子转换比与运行寿期之间关系的分析模型,设计出不同运行寿期的实际堆芯并进行计算,研究了⁶⁰Co产量和中子转换比随高通量工程试验堆（HFETR）运行寿期的变化规律。同时,通过对新燃料元件堆芯的研究找出燃料元件装载量对⁶⁰Co产量和中子转换比的影响,采用点堆模型分析平衡堆芯下HFETR的燃料元件装载量。该优化研究的目的在于为HFETR堆芯装载和运行方式优化提供参考以提高其运行的经济性。结果表明,HFETR运行寿期设计为25 d较佳,在此寿期下的平衡堆芯燃料元件理想装载量为70盒。     

Decay heat production in a TRU burner

The design of new reactors such as ADS has been investigated in many countries during the last years for burning transuranic nuclides (TRUs) contained in spent reactor fuel. To increase the TRU incineration rate, fertile-free dedicated fuels, which may contain a large fraction of minor actinides (MAs), are currently considered. Based on past experience, R&D activities for dedicated fuels in Europe concentrate on fuel forms, in which the oxide actinide phase is mixed with oxide or metal inert matrices. Decay heat in a system with inert matrix fuel (IMF) containing MAs may differ from that in a conventional fast reactor. In this paper, several fast reactor designs with different TRU content are considered and related decay heat values, calculated on the basis JEFF 3.0 and JEFF 3.1 nuclear data libraries, are compared. It is shown that some decay heat components for fuels with MAs may be lower than those for MA-free fuels, but the total decay heat may be significantly higher for cooling times exceeding about 1 min.     

Comparison of sodium and lead-cooled fast reactors regarding reactor physics aspects, severe safety and economical issues

A large number of new fast reactors may be needed earlier than foreseen in the Generation IV plans. According to the median forecast of the Special Report on Emission Scenarios commissioned by the Intergovernmental Panel on Climate Control nuclear power will increase by a factor of four by 2050. The drivers for this expected boost are the increasing energy demand in developing countries, energy security, but also climate concerns. However, staying with a once-through cycle will lead to both a substantially increased amount of high-level nuclear waste and an upward pressure on the price of uranium and even concerns about its availability in the coming decades. Therefore, it appears wise to accelerate the development of fast reactors and efficient re-processing technologies.In this paper, two fast reactor systems are discussed—the sodium-cooled fast reactor, which has already been built and can be further improved, and the lead-cooled fast reactor that could be developed relatively soon. An accelerated development of the latter is possible due to the sizeable experience on lead/bismuth eutectic coolant in Russian Alpha-class submarine reactors and the research efforts on accelerator-driven systems in the EU and other countries.First, comparative calculations on critical masses, fissile enrichments and burn-up swings of mid-sized SFRs and LFRs (600 MW_e) are presented. Monte Carlo transport and burn-up codes were used in the analyses. Moreover, Doppler and coolant temperature and axial fuel expansion reactivity coefficients were also evaluated with MCNP and subsequently used in the European Accident Code-2 to calculate reactivity transients and unprotected Loss-of-Flow (ULOF) and Loss-of-Heat Sink (ULOHS) accidents. Further, ULOFs as well as decay heat removal (protected Total Loss-of-Power, TLOP) were calculated with the STAR-CD CFD code for both systems.We show that LFRs and SFRs can be used both as burners and as self-breeders, homogeneously incinerating minor actinides. The tight pin lattice SFRs (P/D = 1.2) appears to have a better neutron economy than wide channel LFRs (P/D = 1.6), resulting in larger BOL actinide inventories and lower burn-up swings for LFRs. The reactivity burn-up swing of an LFR self-breeder employing BeO moderator pins could be limited to 1.3$ in 1 year. For a 600 MW_e LFR burner, LWR-to-burner support ratio was about two for (U, TRU)O₂-fuelled system, while it increased to approximately 2.8 when (Th, TRU)O₂ fuel was employed. The corresponding figures for an SFR were somewhat lower. The calculations revealed that LFRs have an advantage over SFRs in coping with the investigated severe accident initiators (ULOF, ULOHS, TLOP). The reason is better natural circulation behavior of LFR systems and the much higher boiling temperature of lead. A ULOF accident in an LFR only leads to a 220 K coolant outlet temperature increase whereas for an SFR the coolant may boil. Regarding the economics, the LFR seems to have an advantage since it does not require an intermediate coolant circuit. However, it was also proposed to avoid an intermediate coolant circuit in an SFR by using a supercritical CO₂ Brayton cycle. But in an LFR, the reduced concern about air and water ingress may decrease its cost further.     

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.     

Impact of PWR spent fuel variations on back-end features of advanced fuel cycles with tru-fueled VHTR

Alternative strategies are being considered as management option for current spent nuclear fuel transuranics (TRU) inventory. Creation of transmutation fuels containing TRU for use in thermal and fast reactors is one of the viable strategies. Utilization of these advanced fuels will result in transmutation and incineration of the TRU. The objective of this study is to analyze the impact of conventional PWR spent fuel variations on TRU-fueled very high temperature reactor (VHTR) systems. The current effort is focused on prismatic core configuration operated under a single batch once-through fuel cycle option. IAEA’s nuclear fuel cycle simulation system (VISTA) was used to determine potential PWR spent fuel compositions. Additional composition was determined from the analysis of United States legacy spent fuel that is given in the Yucca Mountain Safety Assessment Report. A detailed whole-core 3-D model of the prismatic VHTR was developed using SCALE5.1 code system. The fuel assembly block model was based on Japan’s HTTR fuel block configuration. To establish a reference reactor system, calculations for LEU-fueled VHTR were performed and the results were used as the basis for comparative studies of the TRU-fueled systems. The LEU fuel is uranium oxide at 15% ²³⁵U enrichment. The results showed that the single-batch core lifetimes ranged between 5 and 7 years for all TRU fuels (3 years in LEU), providing prolonged operation on a single batch fuel loading. Transmutation efficiencies ranged between 19% and 27% for TRU-based fuels (13% in LEU). Total TRU material contents for disposal ranged between 730 and 808 kg per metric ton of initial heavy metal loading, reducing TRU inventory mass by as much as 27%. Decay heat and source terms of the discharged fuel were also calculated as part of the spent fuel disposal consideration. The results indicated strong potential of TRU-based fuel in VHTR.