Characterization of a sodium-cooled fast reactor in an MHR–SFR synergy for TRU transmutation

In the task of destroying the light water reactor (LWR) transuranics (TRUs), we consider the concept of a synergistic combination of a deep-burn (DB) gas-cooled reactor followed by a sodium-cooled fast reactor (SFR), as an alternative way to the direct feeding of the LWR TRUs to the SFR. In the synergy concept, TRUs from LWR are first deeply incinerated in a graphite-moderated DB-MHR (modular helium reactor) and then the spent fuels of DB-MHR are recycled into the closed-cycle SFR. The DB-MHR core is 100% TRU-loaded and a deep-burning (50–65%) is achieved in a safe manner (as discussed in our previous work). In this analysis, the SFR fuel cycle is closed with a pyro-processing technology to minimize the waste stream to a final repository. Neutronic characteristics of the SFR core in the MHR–SFR synergy have been evaluated from the core physics point of view. Also, we have compared core characteristics of the synergy SFR with those of a stand-alone SFR transuranic burner. For a consistent comparison, the two SFRs are designed to have the same TRU consumption rate of ∼250 kg/GW EFPY that corresponds to the TRU discharge rate from three 600 MW DB-MHRs. The results of our work show that the synergy SFR, fed with TRUs from DB-MHR, has a much smaller burnup reactivity swing, a slightly greater delayed neutron fraction (both positive features) but also a higher sodium void worth and a less negative Doppler coefficients than the conventional SFR, fed with TRUs directly from the LWRs. In addition, several design measures have been considered to reduce the sodium void worth in the synergy SFR core.     

Multi-recycling in the ENHS and the equilibrium core

This work investigates the effect of initial fuel composition, power density and number of recycles on the pitch-to-diameter (P/D) ratio and TRans-Uranium isotopes (TRU) loading required for attaining one of the most important design goals of the Encapsulated Nuclear Heat Source (ENHS) – nearly zero burnup reactivity swing over the 20 years of core life. It is found that the required P/D ratio is sensitive to, primarily, the initial concentration of the short-lived isotope ²⁴¹Pu in the fuel loaded into the first core and to the core power density. The longer is the cooling time of the TRU from LWR spent fuel the smaller becomes the relative ²⁴¹Pu concentration and the smaller becomes the fraction of ²⁴¹Pu lost via radioactive decay and, hence, the smaller needs be the conversion ratio required for nearly zero burnup reactivity swing and the larger can be the P/D ratio. Likewise, the higher is the ENHS power density, the smaller becomes the fraction of ²⁴¹Pu lost via radioactive decay and the larger becomes the P/D required for the first core. The optimal P/D ratio tends to increase with the number of times the fuel is recycled from one ENHS core to the next one. The optimal P/D ratio for the equilibrium composition core is in between 1.53 and 1.59. For a given discharge burnup it tends to somewhat increase with the equilibrium core power density. However, if structural materials will be developed to enable a 20 years core life at elevated power densities, the higher the power density the smaller is the required equilibrium P/D ratio.     

Sensitivity Study of Fuel Cost in Extended Burnup BWR Core

A sensitivity study on the fuel cost of an extended burnup BWR core has been carried out on the basis of a realistic model of discharge burnup extension. Full power operating length in months in a refueling cycle and the number of refueling batches are chosen as independent variables in the model to describe extended burnup cores of various types. The reference core for the sensitivity study adopts 9-month full power operation and 4-batch refueling scheme. The difference in the plant cost between the extended burnup core and the reference core, which is referred to as plant capacity factor (PCF) credit, is estimated and combined with the fuel cost to calculate the fuel cost with PCF credit.

The results show that the fuel cost with PCF credit decreases for the extended burnup core with stretched operating length as the burnup extends in cases of constant non-operating length in a cycle, and that it may increase for the extended burnup core with decreased batch number in cases of constant plant capacity factor. It is also suggested that the cost minimum combination of the independent variables can be found to a given discharge burnup for the extended burnup core with decreased batch number in an intermediate case between these two extreme cases. Extended burnup cores with fixed batch number tend to have a lower natural uranium requirement, but larger separative work requirement.

The economic break-even condition for the extended burnup core with decreased batch number is discussed based on the fraction of fixed part in the non-operating length, which is insensitive to the cycle length stretch-out.     

Optimization of seed-blanket type fuel assembly for reduced-moderation water reactor

Parametric studies have been performed for a PWR-type reduced-moderation water reactor (RMWR) with the seed-blanket type fuel assembles to achieve a high conversion ratio, negative void reactivity coefficient and a high burnup by using MOX fuel. From the viewpoint of reactor safety analysis, the fuel temperature coefficients were also studied.From the result of the burnup calculation, it has been seen that ratio of 40–50% of outer blanket in a seed-blanket assembly gives higher conversion ratio compared to the other combination of seed-blanket assembly. And the recommended number of (seed+blanket) layers is 20, in which the number of seed (S) layers is 15 (S15) and blanket (B) layers is 5 (B5). It was found that the conversion ratio of seed-blanket assembly decreases, when they are arranged looks like a flower shape (Hanagara).By the optimization of different parameters, S15B5 fuel assembly with the height of seed of 1000 mm×2, internal blanket of 150 mm and axial blanket of 400 mm×2 is recommended for a reactor of high conversion ratio. In this assembly, the gap of seed fuel rod is 1.0 mm and blanket fuel rod is 0.4 mm. In S15B5 assembly, the conversion ratio is 1.0 and the burnup is 38.18 GWd/t in (seed+internalblanket+outer blanket) region. However, the burnup is 57.45 GWd/t in (seed+internal blanket) region. The cycle length of the core is 16.46 effective full power in month (EFPM) by six batches and the enrichment of fissile Pu is 14.64 wt.%. The void coefficient is +21.82 pcm/%void, however, it is expected that the void coefficient will be negative if the radial neutron leakage is taken into account in the calculation.It is also possible to use S15B5 fuel assembly as a high burnup reactor 45 GWd/t in (seed+internal blanket+outer blanket) region, however, it is necessary to decrease the height of seed to 500 mm×2 to improve the void coefficient. In this reactor, the conversion ratio is 0.97 and void coefficient is +20.81 pcm/%void. The fuel temperature coefficient is negative for both of cases.     

3D in-core fuel management optimization for breed-and-burn reactors

Breed-and-burn (B&B) reactors are a special class of fast reactors that are designed to utilize low grade fuel such as depleted uranium without fuel reprocessing. One of the most challenging practical design feasibility issues faced by B&B reactors is the high level of radiation damage their fuel cladding has to withstand in order to sustain the B&B mode of operation – more than twice the maximum radiation damage cladding materials were exposed to so far in fast reactors. This study explores the possibility of reducing the minimum required peak radiation damage by employment of 3-dimensional (3D) fuel shuffling that enables a significant reduction in the peak-to-average axial burnup, that is, more uniform fuel utilization. A new conceptual design of a B&B core made of axially segmented fuel assemblies was adopted to facilitate the 3D shuffling. Also developed is a Simulated Annealing (SA) algorithm to automate the search for the optimal 3D shuffling pattern (SP). The primary objective of the SA optimization is to minimize the peak radiation damage while its secondary objective is to minimize the burnup reactivity swing, radial power peaking factor and maximum change of fuel assembly power over the cycle. Also studied is the sensitivity of the 3D shuffled core performance to the number of axially stacked sub-assemblies, core height and power level.It was found that compared with the optimal 2-dimensional (2D) shuffled core, the optimal 3D shuffled B&B core made of four 70 cm long axially stacked sub-assemblies and 12 radial shuffling batches offers a 1/3 reduction of the peak radiation damage level – from 534 down to 351 displacements per atom (dpa), along with a 45% increase in the average fuel discharge burnup, and hence, the depleted uranium utilization, while satisfying all major neutronics and thermal-hydraulics design constraints. For the same peak dpa level, the 3D shuffling offers more than double the uranium utilization and the cycle length relative to 2D shuffling. The minimum peak radiation damage is increased to 360 or to 403 dpa if the core is made of, respectively, three – 70 cm or two – 140 cm long axially stacked subassemblies. Reducing the length of the subassemblies of B&B cores made of three-segment assemblies from 70 cm to 60 or 50 cm results in an increase in the peak radiation damage from 360 dpa to, respectively, 368 and 397 dpa.     

TRU self-recycling in a high temperature modular helium reactor

The Deep Burn Project is developing high burnup fuel based on Ceramically Coated (TRISO) particles, for use in the management of spent fuel Transuranics. This paper evaluates the TRU deep-burn in a High Temperature Reactor (HTR) that recycles its own transuranic production. The DB-HTR is loaded with standard LEU fresh fuel and the self-generated TRUs are recycled into the same core (after reprocessing of the original spent fuel). This mode of operation is called self-recycling (SR-HTR). The final spent fuel of the SR-HTR can be disposed of in a final repository, or recycled again.In this study, a single recycling of the self-generated TRUs is considered. The UO₂ fuel kernel is 12% uranium enrichment and the diameter of the kernel is 500 μm. TRISO packing fraction of UO₂ fuel compact is 26%. In the SR-HTR fuel cycle, it is assumed that the spent UO₂ fuel is reprocessed with conventional technology and the recovered TRUs are fabricated into Deep Burn TRISO fuel. The diameter of 200 μm is used for the TRU fuel kernel. A typical coating thickness is used. The core performance is evaluated for an equilibrium cycle, which is obtained by cycle-wise depletion calculations. From the analysis results, the equilibrium cycle lengths of Case 1 (5-ring fuel block SR-HTR) and Case 2 (4-ring fuel block SR-HTR) are 487 and 450 EFPDs (effective full power days), respectively. And the UO₂ fuel discharge burnups of Case 1 and Case 2 are 10.3% and 10.1%, respectively. Also, the TRU discharge burnups of Case 1 and Case 2 are 64.7% and 63.5%, respectively, which is considered extremely high. The fissile (Pu-239 and Pu-241) content of the self-generated TRU vector is about 52%. The deep-burning of TRU in SR-HTR is partly due to the efficient conversion of Pu-240 to Pu-241, which is boosted by the uranium fuel in SR-HTR. It is also observed that the power distribution is quite flat within the uranium fuel zone. The lower power density in TRU fuel is because the TRU burnup is very high. Also, it is found that transmutation of Pu-239 is near complete in SR-HTR and that of Pu-241 is extremely high in all cases. The decay heat of the SR-HTR core is very similar to the UO₂-only core. However, accumulation of the minor actinides is not avoidable in the SR-HTR core. The extreme high burnup of the Deep Burn fuel greatly reduces the amount of heat producing isotopes that could be problematic in spent fuel repositories (like Pu-238).     

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。     

Prospects for Using Mixed Oxide Fuel in Fast Sodium-Cooled Reactors

When designing new fast reactors, it is desirable to increase as much as possible the breeding occurring in the core in order to ensure the minimum excess reactivity for burnup on the one hand and a closed fuel cycle without replenishment with external plutonium and without separating plutonium from uranium during chemical reprocessing of irradiated fuel on the other. The latter requirement greatly decreases the risk of plutonium proliferation in such a fuel cycle. This requires a core breeding ratio 1.05–1.08. Such values can be achieved by using technologically perfected and tested oxide fuel with its volume fraction in the core increased to 55–60%. The results of computational-theoretical studies on the selection and optimization of cores with high fuel fractions for BN-1600 and BN-800 reactors are presented in this article. It is shown that such cores can be built in principle.     

Blanket design studies for maximizing the discharge burnup of liquid metal cooled ATW systems

This paper presents the results of neutronic design studies of lead–bismuth eutectic (LBE) and sodium cooled accelerator transmutation of waste (ATW) blankets. These studies have focused primarily on maximizing the discharge burnup under key thermal-hydraulic and material-related design constraints. Subject to the design constraints on the peak linear power, the maximum coolant velocity, the maximum volume fraction of transuranic (TRU) elements in the dispersion fuel, and the peak fast fluence, design studies have been performed for 840 MW ATW blankets. From these studies, it has been found that the unconstrained discharge burnup for a fixed fuel residence time increases monotonically as the fuel volume fraction and blanket size decrease. The results also show that the discharge burnup is proportional to the peak fast fluence. These indicate that the maximum discharge burnup is primarily determined by imposed design constraints. The maximum discharge burnup achievable under the peak fast fluence limit has been found to be ∼28% for the LBE system, and ∼30% for the sodium system. The optimum fuel volume fraction appears to be ∼0.21 and ∼0.32 for LBE and sodium systems, respectively.     

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.     

Molten salt cooled Encapsulated Nuclear Heat Source (ENHS)-like reactors

The feasibility of designing molten-salt cooled ENHS (Encapsulated Nuclear Heat Source)-like reactor cores with Pu¹⁵N-U¹⁵N nitride fuel for high temperature applications is assessed. The cores considered have uniform fuel composition and no blanket elements and solid reflectors. They are to operate for at least 20 effective full power years without refueling, without fuel shuffling and with burnup reactivity swing less than 0.52%. Three molten-fluoride-salts: NaF(57)-BeF₂(43), ⁷LiF(66)-BeF₂(34), and LiF(46.5)-NaF(11.5)-KF(42) are considered as the coolant and six materials: SS304, Hastelloy-N, HT-9, Mn-316SS, PCA, and SiC, are considered for the structures. It is found that, neutronically, ENHS-like cores can be designed for all combinations of molten-salt coolants and structural materials considered. Relative to the reference ENHS core, the molten-salt cooled cores require significantly tighter lattice, have softer neutron spectra, significantly more negative Doppler reactivity effect, much more positive coolant temperature and void reactivity effect and smaller reactivity worth of the control elements. Of the molten salts considered, LiF-NaF-KF offers the largest p/d ratio and is most suitable for natural circulation cooling.     

Accuracy of nuclear design of fast and thermal neutron coupled core by SRAC

Supercritical-pressure light water cooled fast reactor adopts the blanket fuel assemblies with depleted uranium fuel and zirconium hydride layer in the core for negative coolant void reactivity. Thermal neutrons are generated in the core of fast reactor. It is called “fast and thermal neutron coupled core”. The purpose of the present study is to examine the accuracy of assembly and core calculations including preparation of the macroscopic cross sections with the SRAC code system for “fast and thermal neutron coupled core” in comparison with the Monte Carlo codes, MVP and MVP-BURN. Accuracy of the neutron multiplication factor and coolant void reactivity calculation has been evaluated in four types of cores of different fractions of the blanket fuel assembly with zirconium hydride rods. The conventional analysis is based on the macroscopic cross sections which are prepared with infinite lattice. The conventional SRAC calculation underestimates the neuron multiplication factor for all types of cores. Other findings are that the conventional SRAC calculation overestimates coolant void reactivity for the cores without zirconium hydride rods, and underestimates coolant void reactivity for the core of all blanket fuel assemblies having zirconium hydride rods. To overcome these problems, it has been proposed that the macroscopic cross sections of seed fuel assembly are prepared with the model that a seed fuel assembly is surrounded by blanket fuel assemblies in order to take into account the effects of the surrounding fuel assemblies. Evaluations show that accuracy of the neutron multiplication factor by the SRAC calculation can be improved by the proposed method.     

Effect of Radial Peaking Factor Limitation on Discharge Burnup

Effect of the radial peaking factor limitation on the discharge burnup was examined. In general, lower limitation of the radial peaking factor places restrictions on feasible loading patterns and decreases core performance and economic efficiency. In this paper, relationship between limitation of the radial peaking factor and the discharge burnup was quantitatively investigated in 2-loop and 3-loop PWRs for several cycle lengths and fuel types. Equilibrium cores were generated assuming various radial peaking factor limitations and the change in discharge burnup, which can be considered an index of fuel cycle costs, was evaluated for each case. In order to make accurate comparisons, the generated equilibrium cores were optimized using the OPAL code by the simulated annealing method. From the calculation results, it was revealed that the limitation of the radial peaking factor has considerable impact on the discharge burnup. Relationship between the prediction accuracy of the radial peaking factor and the fuel cycle cost can be also quantitatively estimated from the above results. Therefore, the results can provide a strong motivation to improve in-core fuel management methods.     

Long-Cycle and High-Burnup Fuel Assembly for the VHTR

For a prismatic VHTR fuel assembly, a physics study has been performed to maximize the fuel performance in terms of the cycle length and the discharge burnup for a given fuel enrichment. The relationship between the fuel performance and the fuel configurations has been investigated in terms of the TRISO packing fraction, diameter of the fuel kernel, fuel management, and moderating power of the fuel block. Both a typical low-enrichment uranium fuel (LEU) and a fuel made of transuranics (TRU) from LWR spent fuel are considered in this paper. It is shown that in order to obtain a long refueling cycle and a high burnup at the same time, the fuel loading needs to be increased together with the moderating power of the fuel block. Three ways are considered for a higher moderation of the fuel block: a larger pitch of the coolant hole pattern, an extra graphite thickness in the fuel block, and a higher graphite density. The impact of the increased pitch on the fuel temperature is also evaluated with a thermal analysis code. We have shown that long refueling cycles and high burnups can be achieved simultaneously for both LEU and TRU fuels.     

Effect of TRU fuel loading on core performance and plutonium production of FBR

The effect of trans-uranium (TRU) fuel loading on the reactor core performances as well as the actinide and isotopic plutonium compositions in the core and blanket regions has been analyzed based on the large FBR type. Isotopic plutonium composition of TRU fuel is less than that of MOX fuel except for Pu-238 composition which obtains relatively higher composition. A significant increase of plutonium vector composition is shown by Pu-238 for TRU fuel in the core region as well as its increasing value in the blanket region for doping MA case. Excess reactivity can be reduced significantly (5% at beginning of cycle) and an additional breeding gain can be obtained by TRU fuel in comparison with MOX fuel. Doping MA in the blanket regions reduces the criticality for a small reduction value (0.1%) and it gives a reduction value of breeding ratio. Loading MA in the core regions as TRU fuel composition gives relatively bigger effect to increase the void reactivity coefficient mean while it gives less effect for loading MA in the blanket regions. Similar to the void reactivity coefficient profile, loading MA is more effective to the change of Doppler coefficient in the core regions in comparison with loading MA in the blanket regions which gives slightly less negative Doppler coefficient. Obtained Pu-240 vector compositions in the core region are categorized as practically unusable composition for nuclear device based on the Pellaud's criterion. Less than 7% Pu-240 vector compositions in the blanket region are categorized as weapon grade composition for no doping MA case. Obtaining 9% of Pu-238 composition by doping MA 2% in the blanket regions is enough to increase the level of proliferation resistance for denaturing plutonium based on the Kessler's criterion.     

Uranium-based TRU multi-recycling with thermal neutron HTGR to reduce environmental burden and threat of nuclear proliferation

To reduce environmental burden and threat of nuclear proliferation, multi-recycling fuel cycle with high temperature gas-cooled reactor has been investigated. Those problems are solved by incinerating trans-uranium (TRU) nuclides, which is composed of plutonium and minor actinoid, and there is concept to realize TRU incineration by multi-recycling with fast breeder reactor. In this study, multi-recycling is realized even with a thermal reactor by feeding fissile uranium from outside of the fuel cycle instead of breeding fissile nuclide. In this fuel cycle, recovered uranium and natural uranium are enriched and mixed with recovered TRU to fabricate fresh fuels.

The fuel cycle was designed for a gas turbine high temperature reactor (GTHTR300). Reprocessing is assumed as existing reprocessing with four-group partitioning technology.

As a result, the TRU nuclides excluding neptunium can be recycled by the proposed cycle. The duration of potential toxicity decaying to natural uranium level can be reduced to approximately 300 years, and the footprint of repository for high-level waste can be reduced by 99.7% compared with the standard case. Surplus plutonium is not generated by this cycle. Moreover, incineration of TRU from light water reactor cycle can be performed in this cycle.     

Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of ²³⁹Pu and ²⁴¹Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.     

Neutronics and burnup analysis of an accelerator-based tru-nitride fuel transmutation system with the ATRAS code

The neutronics and burnup analyses of an accelerator-based transmutation system with tungsten target and TRU-nitride fuel were performed with a newly developed code system named ATRAS (Accelerator-based Transmutation Reactor Analysis System). The ATRAS code is an integrated code system which can perform the hadronic cascade process above 20 MeV and neutron transport and core burnup process below 20 MeV with the spallation neutron source.

The specifications of the transmutation system are investigated. The core consists of the central spallation target region and the surrounding TRU-mononitride fuel region. The core is driven by protons at an energy of 1.0 GeV. This system was also proposed as a benchmark problem in the “OECD NEA/NSC Benchmark on Physics aspects of Different Transmutation Concepts”.

According to the calculation results by the ATRAS code, higher power density and transmutation rate were achieved with nitride fuel, and the neutron spectrum was slightly harder than that of the metallic fuel system. The burnup calculation for thermal power 800 MW was also performed with the ATRAS code. It is shown that about 300 kg of TRU are transmuted annually.     

Core design studies for a 1000 MWth Advanced Burner Reactor

This paper describes the core design and performance characteristics of 1000 MW_th Advanced Burner Reactor (ABR) core concepts with a wide range of TRU conversion ratio. Using ternary metal alloy and mixed oxide fuels, reference core designs of a medium TRU conversion ratio of ∼0.7 were developed by trade-off between burnup reactivity loss and TRU conversion ratio. Based on these reference core concepts, TRU burner cores with low and high TRU conversion ratios were developed by changing the intra-assembly design parameters and core configurations. Reactor performance characteristics were evaluated in detail, including equilibrium cycle core performances, reactivity feedback coefficients, and shutdown margins. The results showed that by employing different assembly designs, a wide range of TRU conversion ratios from ∼0.2 to break-even can be achieved within the same core without introducing significant performance and safety penalties.     

Advanced fuel cycle scenarios for High Temperature Gas Reactors

This study evaluates nuclear fuel cycle scenarios which are based on recycling spent nuclear fuel for the sustainability of nuclear energy. Three fuel cycle scenarios, the Light Water Reactor (LWR)–Advanced Recycling Reactor (ARR) recycle, the LWR–High Temperature Gas Reactor (HTGR)–ARR recycle, and the HTGR partial recycling fuel cycle, are assessed for their mass flow and electricity generation costs and the results are compared to those of the LWR once-through fuel cycle. The spent fuels are recycled in both the Consolidated Fuel Treatment Center and the Actinide Management Island, which are capable of reprocessing spent fuels by Uranium Extraction and Pyrochemical processes, respectively. The mass flow calculations show that the Transuranics (TRU) which have a long-term radiation effect can be completely burned in the recycling fuel cycles, resulting in 350, 450 and 6 times reduction of TRU inventory for the LWR–ARR, LWR–HTGR–ARR and HTGR partial recycling fuel cycles, respectively, when compared to the once-through fuel cycle. The electricity generation costs of these fuel cycle scenarios were estimated to be 39.1, 34.9 and 25.7 USD/MW h(e), which are comparable to or smaller than that of the once-through fuel cycle. Although the candidate fuel cycles adopt reprocessing options which raise fuel cycle cost, increase in uranium cost and the advanced design of the HTGR can further reduce the advanced fuel cycle costs of the HTGR.