Neutronics analysis of megawatt-class gas-cooled space nuclear reactor design

ABSTRACT

Neutronics analysis was conducted for a proposed megawatt-class gas cooled space nuclear reactor design. The reactor design has a high operating temperature of up to 1500 K. Annular UO₂ fuel rods were used to reduce the central temperature of the fuel. The thermal power is 2.3 MWt and is converted into electric power by a direct Brayton cycle. The control rods were arranged in different configurations and were analyzed in order to evaluate the influence on the reactor design in different scenarios. The calculation results reveal that the control rods arrangements have influences on the begin-of-life (BOL) excess reactivity and the shutdown reactivity. The distribution of control rods affects the neutron economy and leakage in the fuel region, consequently affecting the reactivity. It is also known that the reactivity in flooding scenarios are not sensitive to different control rod arrangements. Meanwhile, according to calculation results, the proposed reactor design has enough shutdown reactivity margin which will allow for flexible control strategy. Further analysis is still needed for more detailed and accurate parameters of the reactor design.     

Whole Core Calculations of Power Reactors by Use of Monte Carlo Method

Abstract

Whole core calculations have been performed for a commercial size PWR and a prototype LMFBR by using vectorized Monte Carlo codes. Geometries of cores were precisely represented in a pin by pin model. The calculated parameters were k _eff, control rod worth, power distribution and so on. Both multigroup and continuous energy models were used and the accuracy of multigroup approximation was evaluated through the comparison of both results. One million neutron histories were tracked to considerably reduce variances. It was demonstrated that the high speed vectorized codes could calculate k _eff, assembly power and some reactivity worths within practical computation time. For pin power and small reactivity worth calculations, the order of 10 million histories would be necessary. It would be difficult for the conventional scalar code to solve such large scale problems while the present codes consumed computation time less than 30 min for a PWR and 1 hour for an LMFBR. Required number of histories to achieve target design accuracy were estimated for those neutronic parameters.     

Benchmark Problem Suite for Reactor Physics Study of LWR Next Generation Fuels

This paper proposes a benchmark problem suite for studying the physics of next-generation fuels of light water reactors. The target discharge burnup of the next-generation fuel was set to 70GWd/t considering the increasing trend in discharge burnup of light water reactor fuels. The UO₂ and MOX fuels are included in the benchmark specifications. The benchmark problem consists of three different geometries: fuel pin cell, PWR fuel assembly and BWR fuel assembly. In the pin cell problem, detailed nuclear characteristics such as burnup dependence of nuclide-wise reactivity were included in the required calculation results to facilitate the study of reactor physics. In the assembly benchmark problems, important parameters for in-core fuel management such as local peaking factors and reactivity coefficients were included in the required results. The benchmark problems provide comprehensive test problems for next-generation light water reactor fuels with extended high burnup. Furthermore, since the pin cell, the PWR assembly and the BWR assembly problems are independent, analyses of the entire benchmark suite is not necessary: e.g., the set of pin cell and PWR fuel assembly problems will be suitable for those in charge of PWR in-core fuel management, and the set of pin cell and BWR fuel assembly problems for those in charge of BWR in-core fuel management.     

Conceptual design of a small nuclear reactor for large-diameter NTD-Si using short PWR fuel assemblies

A design of a small nuclear reactor for a large-diameter NTD-Si using a conventional Pressurized Water Reactors (PWR) full-length assembly was proposed in previous works. The height of the full-length assembly was 400 cm, and the overall size of the reactor and reflector around the core became large. In addition, the irradiation channel became very long, making handling of the Si ingots in the channel more difficult. The use of a short PWR fuel assembly, with a height of 100 cm, was considered in the current work. With the shorter assembly, the design of the reactor became compact and more practical. Gd₂O₃ and control rods were used to suppress excess reactivity. Criticality, neutron transport, and core burn-up calculations were performed using the MVP/GMVP II code and MVP-BURN code. Steady-state single-channel thermal hydraulic analyses were also performed. The calculation results showed that the reactor could be critical over 1200 days, and that heat removal from core was possible under 1 atm operating pressure. Large-diameter ingot up to 20 cm in height could be doped with sufficient uniformity. The reactor semiconductor production rate was estimated, and varied between 48 tons/year and 70 tons/year for the 50 Ω cm target resistivity depending on the position of the control rod.     

Technetium transmutation in thin layer coating on PWR fuel rods

Based on the EFTTRA-T2 experiment results, we study the transmutation characteristics of pressurized water reactors (PWR) after coating a thin layer of Tc-99 on the fuel rods. Our calculation shows that for the same Tc-99 loading amount, the effect on the PWR k_eff after coating Tc-99 on the PWR fuel rods is much less than that of the homogeneous addition of Tc-99 to uranium dioxide nuclear fuel. If we just coat 0.2λ_c (0.0065 mm) thickness Tc-99 on PWR fuel rods, the total Tc-99 coating amount is about 291.37 kg, this is approximately equivalent to the 4 PWR Tc-99 annual outputs, and the system k_eff merely decreases to 0.98530.Loading Tc-99 to the PWR is equivalent to introducing extra poisons to PWR system to control excess reactivity, some control poisons like boric acid concentration in primary coolant or burnable poison rods in fuel assemblies are needed to be removed to keep the reactor in criticality. As Tc-99 coating thickness increases from 0.05λ_c to 0.2λ_c, no matter which substitution pattern is used, B16→12 or C16→12, the system k_eff variations are almost the same and can return to criticality again after removing corresponding burnable poison rods from fuel assemblies. For coating 0.15λ_c or 0.2λ_c thickness on the fuel rods of PWR, the system k_eff is slightly below the criticality either in B16→12 or C16→12 substitution pattern, we may reduce the concentration of the boric acid slightly to let the system in criticality again.Our calculation results indicate that the optimal coating thickness of Tc-99 on PWR fuel rods is probably between 0.15λ_c to 0.2λ_c, i.e. 0.00488–0.0065 mm.     

Neutronic analysis of hydride fueled PWR cores

The neutronic properties of U-ZrH_1.6 fuelled PWR cores are investigated and compared against those of the currently used UO₂ fuelled cores. In the first part of this work a parametric study is performed to quantify the neutronically achievable burnup for both hydride and oxide fuels at a number of enrichment levels and for a large number of geometries covering a wide design space of fuel rod outer diameter, D, and lattice pitch, P. The fuel temperature and coolant temperature reactivity coefficients as well as the small and large void reactivity coefficients are calculated for hydride fuel with 5% and 12.5% enriched uranium. For this purpose a simplified procedure was developed that can, using single unit cell or assembly calculations, (1) account for non-linear burnup dependent k_∞ and thus to adequately predict the discharge burnup; (2) estimate the burnup dependent soluble boron concentration and; (3) estimate the reactivity coefficients; all of the above for a multi-batch core. In the second part of this work a detailed neutronic analysis is carried out for the six most economical geometries of both oxide and hydride fuels, with the purpose of designing the U-ZrH_1.6 fueled PWR cores to have negative reactivity coefficients. The preferred design found is replacement of 25 ^v/_o of the ZrH_1.6 by thorium hydride, along with addition of some IFBA burnable poison. It is also found that the conversion from oxide to hydride fueled PWR cores could be done without modifications in the control system.     

Design simplification of a small nuclear reactor for large-diameter neutron transmutation doping silicon using control rods

A design concept for a small nuclear reactor for neutron transmutation doping silicon (NTD-Si) using a Pressurized Water Reactor (PWR) full-length fuel assembly was proposed in our previous work. The excess reactivity was suppressed by a combination of Gd₂O₃ and soluble boron, which results in a flatter flux profile over the core than with control rod insertion; however, the soluble boron system for reactivity control is quite complex and expensive. The removal of this system would make the design much simpler. In the current work, the removal of soluble boron is considered. Criticality, neutron transportation and core burn-up calculations were performed using the MVP/GMVP II code and MVP-BURN code. The calculation results show that the insertion of control rods in five of the nine assemblies is enough to suppress reactivity. The thermal hydraulic analysis showed that heat removal from the core was possible under 1 atm operating pressure. Silicon ingots up to 30 cm in diameter could be irradiated with sufficient uniformity in the irradiation channels.     

A small long-cycle PWR core design concept using fully ceramic micro-encapsulated (FCM) and UO2–ThO2 fuels for burning of TRU

In this paper, a new small pressurized water reactor (PWR) core design concept using fully ceramic micro-encapsulated (FCM) particle fuels and UO₂–ThO₂ fuels was studied for effective burning of transuranics from a view point of core neutronics. The core of this concept rate is 100 MWe. The core designs use the current PWR-proven technologies except for a mixed use of the FCM and UO₂–ThO₂ fuel pins of low-enriched uranium. The significant burning of TRU is achieved with tri-isotropic particle fuels of FCM fuel pins, and the ThO₂–UO₂ fuel pins are employed to achieve long-cycle length of ～4 EFPYs (effective full-power year). Also, the effects of several candidate materials for reflector are analyzed in terms of core neutronics because the small core size leads to high sensitivity of reflector material on the cycle length. The final cores having 10 w/o SS303 and 90 w/o graphite reflector are shown to have high TRU burning rates of 33%–35% in FCM pins and significant net burning rates of 24%–25% in the total core with negative reactivity coefficients, low power peaking factors, and sufficient shutdown margins of control rods.     

Transient Xenon Analysis in a Molten Salt Breeder Reactor

An evaluation is made to estimate the transient xenon behavior in an MSBR for several representative patterns of operation. Such analysis is indispensable for detailed evaluation of reactivity balance under transient conditions. The results are compared with those of a typical PWR. The xenon behavior does not differ between the two types of reactor to the extent that might be expected from the fact that in the MSBR, xenon behavior is additionally conditioned by the processes of migration into the circulating bubbles and into the graphite, as well as by diffusion therein.

It is shown that the reactivity transients due to xenon buildup can be held within the range of counteraction by control rod movement for any normal change of reactor output, so long as the reactor is not shut down. After a shutdown, insertion of the control rods will not suffice to override the xenon buildup, but then the fuel processing system could be conveniently utilized to increase the quantity of ²³³U contained in the fuel and regain required reactivity of the core.     

Economic analysis of grid and wire wrap supported hydride and oxide fueled pressurized water reactors

An economic analysis is performed to calculate the levelized unit cost of electricity (COE) for a pressurized water reactor (PWR) retrofitted with a range of potential U (45 wt.%)-ZrH_1.6 hydride and UO₂ oxide fueled geometries (i.e., combinations of rod diameter and pitch) supported by traditional grid spacers (square array) and wire wrap spacers (hexagonal array). The time frame considered in computing the COE is the remaining plant life, beginning at the time of retrofit. The goals of the analysis are twofold: (1) comparing the economic performance of UO₂ and U-ZrH_1.6 fuels for a range of retrofitted geometries supported by grid and wire wrap spacers; and (2) investigating the potential economic benefits for nuclear utilities considering retrofitting new fuels and/or geometries into existing PWR pressure vessels. Fuel cycle, operations and maintenance (O & M), and capital costs are considered.The economic performance of U-ZrH_1.6 and UO₂ fuels is found to be similar, with UO₂ fueled designs providing a slight advantage when supported by grid spacers, and U-ZrH_1.6 providing a slight advantage when supported by wire wrap spacers. These small differences in cost, however, are within the bounds of uncertainty of this study and are not believed to provide a strong economic argument for the use of one fuel type over the other.To demonstrate the potential economic benefits of retrofitted designs to nuclear utilities, two different comparisons are made. The first compares the COE for retrofitted designs with the COE for a reference PWR, assumed to have operated long enough to recuperate its initial capital investment. The costs for this reference PWR reflect the “do-nothing” case for current plant owners whose primary expenditures are fuel cycle and O & M costs. The second comparison introduces a different reference PWR that includes the costs to operate an existing unit and the cost to purchase power from a newly constructed PWR, for comparison with retrofitted designs which offer increased power relative to existing commercial PWRs.For the first comparison, no grid supported designs and only one wire wrap supported design (i.e., U-ZrH_1.6 Stretch Case) provide a lower levelized unit cost of electricity than the reference “do-nothing” PWR. The primary cause of this conclusion is the capital costs incurred by retrofitted designs to change the core geometry and, for many designs, to upgrade primary and secondary loop components for operation at higher power than the reference PWR. The reference “do-nothing” PWR cost in this first comparison includes only operations and maintenance as well as fuel cycle costs but does not include a capital component. For the second comparison, significant cost savings are demonstrated for both grid (15-19% savings) and wire wrap (30-40% savings) supported designs using U-ZrH_1.6 and UO₂ fuels. These cost savings are enabled by enhancing the pumping capacity of the primary system and, for wire wrap supported designs, by taking advantage of enhanced critical heat flux performance. The optimal geometry for retrofitted UO₂ and U-ZrH_1.6 fueled PWR cores supported by grid spacers is D_rod = 6.5 mm and P/D = 1.39. The cost savings over the second case reference PWR are ∼19 and 15%, respectively. The cost savings for retrofitted PWRs that incorporate wire wrap spacing are even larger because of operation at even higher power. Cost savings over the reference PWR range between 30 and 40% for the U-ZrH_1.6 and UO₂ Achievable and Stretch Cases. The optimal geometries for the U-ZrH_1.6 Achievable and Stretch Cases are D_rod = 8.08 mm, P/D = 1.41 and D_rod = 8.71 mm, P/D = 1.39, respectively. The optimal geometries for the UO₂ Achievable and Stretch Cases are D_rod = 7.13 mm, P/D = 1.42 and D_rod = 9.34 mm, P/D = 1.27, respectively. Utilities seeking to meet rising demand by expanding capacity may therefore strongly benefit from retrofitting existing PWRs with either U-ZrH_1.6 or UO₂ fueled designs. These new designs have different geometries than are currently used by commercial plants. A conclusion on which fuel type to use, however, could not be reached in this analysis as both offer similar economic performance.     

Pu recycling in a full Th-MOX PWR core. Part I: Steady state analysis

Current practice of Pu recycling in existing Light Water Reactors (LWRs) in the form of U-Pu mixed oxide fuel (MOX) is not efficient due to continuous Pu production from U-238. The use of Th-Pu mixed oxide (TOX) fuel will considerably improve Pu consumption rates because virtually no new Pu is generated from thorium. In this study, the feasibility of Pu recycling in a typical pressurized water reactor (PWR) fully loaded with TOX fuel is investigated.Detailed 3-dimensional 100% TOX and 100% MOX PWR core designs are developed. The full MOX core is considered for comparison purposes. The design stages included determination of Pu loading required to achieve 18-month fuel cycle assuming three-batch fuel management scheme, selection of poison materials, development of the core loading pattern, optimization of burnable poison loadings, evaluation of critical boron concentration requirements, estimation of reactivity coefficients, core kinetic parameters, and shutdown margin.The performance of the MOX and TOX cores under steady-state condition and during selected reactivity initiated accidents (RIAs) is compared with that of the actual uranium oxide (UOX) PWR core.Part I of this paper describes the full TOX and MOX PWR core designs and reports the results of steady state analysis. The TOX core requires a slightly higher initial Pu loading than the MOX core to achieve the target fuel cycle length. However, the TOX core exhibits superior Pu incineration capabilities.The significantly degraded worth of control materials in Pu cores is partially addressed by the use of enriched soluble boron and B₄C as a control rod absorbing material. Wet annular burnable absorber (WABA) rods are used to flatten radial power distribution. The temperature reactivity coefficients of the TOX core were found to be always negative. The TOX core has a slightly reduced, as compared to UOX core, but still sufficient shutdown margin.In the TOX core β_eff is smaller by about a factor of two in comparison to the UOX core and even lower than that of the MOX core. The combination of small β_eff and reduced control materials worth may potentially deteriorate the performance under RIA conditions and requires an additional examination. The behavior of the considered cores during the most limiting RIAs, such as rod ejection, main steam line break, and boron dilution, is further investigated and reported in Part II of the paper.     

Small PWR Using Coated Particle Fuel of Thorium and Plutonium

An innovative concept of PFPWR50 for district heating has been studied, which is a small PWR of 50MWt capability using coated particle fuels with conventional zircaloy cladding. This concept takes advantages of fuel integrity against fission products release of coated particle fuels and a high reliability of PWR technology based on the long history of a successful operation. We have investigated burnup characteristics of fuel rods, assemblies, and reactor cores by the calculation code SRAC95 in order to establish a core concept of long life without on-site refueling. The loading pattern of assemblies with various concentrations of burnable poison is optimized to obtain a flat excess reactivity during the core life in order to eliminate a soluble boron control system. The core life of a cycle is about 8.9 equivalent full power years. And we have also studied the applicability of SiC/SiC composite cladding in place of zircaloy cladding, which is now under development for gas cooled fast reactor fuels. It could be applicable to high burnup fuel rods for a long term operation. From the calculation results, it is found out that the burnup characteristics do not change significantly with SiC cladding and contribute to elongate the core life to 9.2 equivalent full power years.     

Experimental Results of Some Cluster Tests in NSRR

The NSRR programme is in progress in JAERI using a pulsed reactor to evaluate the behavior of reactor fuels under reactivity accident conditions. This report describes briefly the experimental results and preliminary analysis of two cluster tests.

In the cluster configuration of five fuel rods, the power distribution in outer fuel rods are not symmetric due to neutron absorption in central fuel rod. The cladding temperature on the exterior boundaries of the cluster is higher than that in interior. Good agreement was obtained between the calculated and measured cladding temperature histories. In the 3.8$ excess reactivity test, cluster averaged energy deposition of 237 cal/g-UO₂, cladding melting and deformation were limited to the portions of the fuel rods that were on the exterior boundaries of the cluster.     

Gas-cooled fast breeder reactor designs with advanced fuel and cladding

This paper examines the potential impact of some alternative cladding and fuel materials being considered for the liquid metal fast breeder reactor (LMFBR) on the performance and design of large commercial gas-cooled fast breeder reactors (GCFRs). Mixed carbide fuel and Inconel 718 cladding material were examined. Another cladding alternative considered was silicon carbide (SiC), which presents some interesting possibilities in high-temperature performance. Design concepts based on the above fuel and claddings were examined and compared with a reference oxide/316 stainless steel design based on a commercial 4000 MW(th) [1500 MW(e)] system. Substantial benefits can be derived from a high-temperature cladding material such as Inconel 718 or ¹⁶Pe; core volume and steam generator heat transfer area could be reduced by 20% or more, and significant reductions in core inventory and doubling time are possible. Carbide fuels would reduce the number of fuel rods by 50% because of higher linear power, and doubling time would be lowered.     

压水堆堆芯设计特点及其演变

综合论述了压水堆堆芯设计中的化学补偿反应性、标准化无盒大型燃料组件、棒束型控制棒、可燃毒物和采用多区堆芯装料等基本问题。并以上述５大问题为基础,简要叙述了负荷跟踪运行给堆芯设计带来的有关设计问题。此外,简要介绍了当前压水堆堆芯的改进设计及演变过程。     

三种堆型核电厂经济性评价

本文分析了核电投资的特点、建立了考虑价格浮动和通货膨胀等因素影响的核电厂建成价和核燃料成本的计算模型。对压水堆、高温气冷堆和快堆三种堆型的经济性进行了研究。结果表明，当高温气冷堆和快堆两种先进堆型实现商用概念设计后，其商业竞争能力可与现有的压水堆相媲美。     

Reactivity control options of space nuclear reactors

This paper compares two ex-core control options of the gas-cooled Submersion Subcritical Safe Space (S^4) reactor with a fast neutrons energy spectrum: (a) rotating BeO drums with 120° thin segments of enriched B₄C in the BeO radial reflector; and (b) sliding segments in the BeO radial reflector. Investigated are the effects on the beginning-of-life (BOL) excess reactivity, reactivity depletion rate and operation life, and the spatial neutron flux distributions and fission power profiles in the core. Also investigated is the effect of reducing the thickness of the enriched B₄C segments in the control drums on the BOL excess reactivity, when one or two of the 6 drums are stuck in the shutdown position. Reducing the thickness of the B₄C segments from 0.5 mm to 0.238 mm, with one drums stuck in the shutdown position, increases BOL cold and hot-clean excess reactivity from +$1.71 and +$0.47 to +$2.38 and +$0.89, respectively. These reactivity values are almost identical to those of the reactor with one of the six reflector segments stuck open in the shutdown position. Results also showed that the control options made little difference in the reactor performance. The power peaking in the reactor core with sliding reflector segments is slightly lower and the spatial power profiles are relatively flatter. The operation life of the reactor with a sliding reflector segments control, when operating at a nominal thermal power of 471 kW, is only 22 full power days longer than with rotating drums control.     

Estimation of Spent Fuel Compositions from Light Water Reactors

The compositions and quantities of minor actinide (MA) and fission product (FP) in spent fuels will be diversified with the use of high discharged burnup fuels and MOX fuels in LWRs which will be a main part of power reactors in future.

In order to investigate above diversities, we have studied on the calculation method to be used in the estimation of spent fuel compositions and adopted the real irradiation calculation in which axial burnup and moderator distribution are considered in the burnup calculation.

On the basis of the calculations, compositions and burnup quantities of various LWR spent fuels (reactor type: PWR and BWR, discharged burnup: 33, 45 and 60 GWd/tHM, fuel type: U0₂ and MOX) are apparently estimated among various forms of fuels. As an example, it is shown that there are considerable discrepancy in MA burnup between PWR and BWR spent fuels.     

Heavy Water Zero Power Reactor (HWZPR) mixed core first criticality,calculation and experiments

HWZPR original fuel is natural U metal fuel but other kinds of fuel can also be utilized. In a research work on UO₂ fuel, reactor fuel was partially replaced by natural UO₂ fuel and physical parameters of the new core were compared with original core. Thirty six natural U metal fuel rods were substituted by natural UO₂ fuel assemblies. Prior to the first criticality operation with the new core, it was simulated by stochastic and deterministic calculation methods i.e. MCNP-4C and WIMS-CITATON codes, respectively. In order to investigate criticality and safety of the mixed core, important reactor physics parameters such as effective multiplication factor (K_eff) at different water levels, critical water level, reactivity worth of D₂O and reactivity worth of safety and control rods were calculated.The calculated results ensured reactor criticality and satisfied reactor safety criteria. Therefore, with the permission of the reactor safety committee, the first criticality operation was performed successfully. Later, during a series of reactor operations important physical parameters were measured experimentally. There is good consistency between the theoretical and experimental results.     

Evaluation of high plutonia (44% PuO2) MOX as a fuel for fast breeder test reactor

Uranium plutonium mixed oxide (MOX) containing up to 30% plutonia is the conventional fuel for liquid metal cooled fast breeder reactor (LMFBR). Use of high plutonia (>30%) MOX fuel in LMFBR had been of interest but not pursued. Of late, it has regained importance for faster disposition of plutonium and also for making compact fast reactors. Some of the issues of high plutonia MOX fuels which are of concern are its chemical compatibility with liquid sodium coolant, dimensional stability and low thermal conductivity. Available literature information for MOX fuel is limited to a plutonium content of 30%. Thermodynamic assessment of mixed oxide fuels indicate that with increasing plutonia oxygen potential of the fuel increases and the fuel become more prone to chemical attack by liquid sodium coolant in case of a clad breach. In the present investigation, some of these issues of MOX fuel have been studied to evaluate this fuel for its use in fast reactor. Extensive work on the out-of-pile thermo-physical properties and fuel-coolant chemical compatibility under different simulated reactor conditions has been carried out. Results of these studies were compared with the available literature information on low plutonia MOX fuel and critically analyzed to predict in reactor behaviour of this fuel containing 44% PuO₂. The results of these out-of-pile studies have been very encouraging and helped in arriving at a suitable and achievable fuel specification for utilization of this fuel in fast breeder test reactor (FBTR). As a first step of test pin irradiation programme in FBTR, eight subassemblies of the MOX fuel are undergoing irradiation in FBTR.