Influence of void fraction change on plutonium and minor actinides recycling in BWR with equilibrium burnup

A study on the influence of void fraction change on plutonium and minor actinides recycling in standard boiling water reactor (BWR) with equilibrium burnup model has been conducted. We considered the equilibrium burnup model since it is a simple time independent burnup method that can handle all possible produced nuclides in any nuclear system.

The uranium enrichment for the criticality of the reactor diminishes significantly for the plutonium and minor actinides recycling case compared to that of the once-through cycle of BWR case. This parameter decreases much lower with the increasing of the void fraction. A similar propensity was also shown in the required natural uranium per annum. The annual required natural uranium was calculated by assuming that the uranium concentration in the tail of the enrichment plant is 0.25 w%. The amount of loaded fuel reduces slightly with the increment of the void fraction for plutonium and minor recycling in BWR.     

Study of the radiotoxicity of actinides recycling in boiling water reactors fuel

In this paper the production and destruction, as well as the radiotoxicity of plutonium and minor actinides (MA) obtained from the multi-recycling of boiling water reactors (BWR) fuel are analyzed. A BWR MOX fuel assembly, with uranium (from enrichment tails), plutonium and minor actinides is designed and studied using the HELIOS code. The actinides mass and the radiotoxicity of the spent fuel are compared with those of the once-through or direct cycle. Other type of fuel assembly is also analyzed: an assembly with enriched uranium and minor actinides; without plutonium. For this study, the fuel remains in the reactor for four cycles, where each cycle is 18 months length, with a discharge burnup of 48 MWd/kg. After this time, the fuel is placed in the spent fuel pool to be cooled during 5 years. Afterwards, the fuel is recycled for the next fuel cycle; 2 years are considered for recycle and fuel fabrication. Two recycles are taken into account in this study. Regarding radiotoxicity, results show that in the period from the spent fuel discharge until 1000 years, the highest reduction in the radiotoxicity related to the direct cycle is obtained with a fuel composed of MA and enriched uranium. However, in the period after few thousands of years, the lowest radiotoxicity is obtained using the fuel with plutonium and MA. The reduction in the radiotoxicity of the spent fuel after one or two recycling in a BWR is however very small for the studied MOX assemblies, reaching a maximum reduction factor of 2.     

Design of a boiling water reactor core based on an integrated blanket–seed thorium–uranium concept

This paper is concerned with the design of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material in an integrated blanket–seed (BS) assembly. The integrated BS concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned out in a once-through cycle. The idea behind the lattice design is to use the thorium conversion capability in a BWR spectrum, taking advantage of the ²³³U build-up. A core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average ²³⁵U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the fuel assembly.     

A BWR fuel assembly design for efficient use of plutonium in thorium–plutonium fuel

The objective of this study is to develop an optimized BWR fuel assembly design for thorium–plutonium fuel. In this work, the optimization goal is to maximize the amount of energy that can be extracted from a certain amount of plutonium, while maintaining acceptable values of the neutronic safety parameters such as reactivity coefficients, shutdown margins and power distribution. The factors having the most significant influence on the neutronic properties are the hydrogen-to-heavy-metal ratio, the distribution of the moderator within the fuel assembly, the initial plutonium fraction in the fuel and the radial distribution of the plutonium in the fuel assembly. The study begins with an investigation of how these factors affect the plutonium requirements and the safety parameters. The gathered knowledge is then used to develop and evaluate a fuel assembly design. The main characteristics of this fuel design are improved Pu efficiency, very high fractional Pu burning and neutronic safety parameters compliant with current demands on UOX fuel.     

低泄漏堆芯燃料管理的一种多循环优化方法

提出一种用于指导压水堆低泄漏堆芯燃料管理的多循环优化方法。该方法将多循环优化问题分解为３步优化处理：首先用线性规划确定满足多循环总体目标最优的各个单循环优化目标参数,然后以此为条件,对多循环中相继的各个单循环进行燃料组件的优化布置,最后进行可燃毒物的优化配置。本文着重讨论第一步优化方法,并给出主要计算结果。     

Plutonium breeding of light water cooled fast reactors

Light water cooled fast reactor with new fuel assemblies (FA) has been studied for high breeding of fissile plutonium. It achieves fissile plutonium surviving ratio (FPSR) of 1.342 (discharge/loading), 1.013 end and beginning of equilibrium cycle (EOEC/BOEC), and compound system doubling time (CSDT) of 95.9 years at the average coolant density of pressurized water reactor (PWR). It is further improved for reduced moderation boiling water reactor (BWR) (RMWR) coolant density. Fissile plutonium surviving ratio reaches 1.397 (discharge/loading), 1.030 (EOEC/BOEC) and CSDT is 37 years. The present study has shown the possibility of breeding at the PWR coolant density and meeting the growth rate of energy demand of advanced countries at the RMWR and Super FR coolant density for the first time. The new FA consist of closely packed fuel rods. The integrity of welding of fuel rods at the top and bottom ends is maintained as the conventional fuel rods. The coolant to fuel volume fraction is reduced to 0.085, one-sixth of that of RMWR. The volume fraction remains unchanged with the diameter of the fuel rod. The thermal hydraulic design of the cores remains for the future study.     

Design of an overmoderated fuel and a full MOX core for plutonium consumption in boiling water reactors

The use of uranium–plutonium mixed oxide fuel (MOX) in light water reactors (LWR) is nowadays a current practice in several countries. Generally 1/3 of the reactor core is loaded with MOX fuel assemblies and the other 2/3 with uranium assemblies. Nevertheless the plutonium utilization could be more effective if the full core could be loaded with MOX fuel. In this paper the design of a boiling water reactor (BWR) core fully loaded with an overmoderated MOX fuel design is investigated. The design of overmoderated BWR MOX fuel assemblies based on a 10×10 lattice are developed, these designs improve the neutron spectrum and the plutonium consumption rate, compared with standard MOX assemblies. In order to increase the moderator to fuel ratio two approaches are followed: in the first approach, 8 or 12 fuel rods are replaced by water rods in the 10×10 lattice; in the second approach, an 11×11 lattice with 24 water rods is designed with an active fuel length very close to the standard MOX assembly. The results of the depletion behavior and the main steady state core parameters are presented. The feasibility of a full core loaded with the 11×11 overmoderated MOX fuel assembly is verified. This design take advantage of the softer spectrum comparable to the 10×10 lattice with 12 water rods but with thermal limits comparable to the standard MOX fuel assembly.     

Fuel lattice design with Path Relinking in BWR’s

In this paper a new system, called ohtli-PR, is presented. This system was developed for fuel lattice design optimization in Boiling Water Reactors (BWR) using the Path Relinking and Scatter Search techniques. As input data, the system uses an initial uranium enrichment and gadolinia percentage. At the beginning of the optimization process, the system generates a random seed fuel lattice, which is used to generate a set with 100 fuel lattices. From this set, a subset is chosen and then an iterative process is carried out until an optimized fuel lattice design is attained. During the iterative process, the system performs at each iteration, two stages. In the first one, the Scatter Search technique is applied. In this stage, a new fuel lattice is constructed using combination pairs of elements. In the second stage, two elements are used to create a path between the best and the worst solutions of the initial subset; this stage is known as Path Relinking. In both stages, if the new element does not improve any solution of the subset, the process continues with the same subset in the next iteration, but using another pair of elements to build either a new combination or a new path. The system uses an objective function with both the Power Peaking Factor (PPF) and the Infinite Neutron Multiplication Factor (k_inf) at the beginning of life of the fuel lattice. The main idea is to keep the Infinite Neutron Multiplication Factor inside a proposed interval and to minimize the Power Peaking Factor. These parameters are calculated with the CASMO-4 code. When the PPF and k_inf parameters are satisfied, the fuel lattice is evaluated taking into account a fuel reload and its respective control rod pattern to verify both the energy obtained and the thermal limits. In this work, the designed fuel lattices correspond to the bottom of the fuel assembly and two different uranium enrichments were used.     

Monte Carlo calculations of the incineration of plutonium and minor actinides of laser fusion inertial confinement fusion fission energy(LIFE) engine

In this paper, neutronic analysis in a laser fusion inertial confinement fusion fission energy(LIFE) engine fuelled plutonium and minor actinides using a MCNP codes was investigated.LIFE engine fuel zone contained 10 vol% TRISO particles and 90 vol% natural lithium coolant mixture. TRISO fuel compositions have Mod(1): reactor grade plutonium(RG-Pu), Mod(2):weapon grade plutonium(WG-Pu) and Mod(3): minor actinides(MAs). Tritium breeding ratios(TBR) were computed as 1.52, 1.62 and 1.46 for Mod(1), Mod(2) and Mod(3), respectively. The operation period was computed as ～21 years when the reference TBR??1.05 for a selfsustained reactor for all investigated cases. Blanket energy multiplication values(M) were calculated as 4.18, 4.95 and 3.75 for Mod(1), Mod(2) and Mod(3), respectively. The burnup(BU)values were obtained as ～1230, ～1550 and ～1060 GWd tM~(-1), respectively. As a result, the higher BU were provided with using TRISO particles for all cases in LIFE engine.     

Reactivity Coefficients in BN-600 Core with Minor Actinides

In 1999, the IAEA has initiated a Coordinated Research Project on “Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effects.” Three benchmark models representing different modifications of the BN-600 fast reactor have been sequentially established and analyzed, including a hybrid core with highly enriched uranium oxide and MOX fuel, a full MOX core with weapons-grade plutonium, and a MOX core with plutonium and minor actinides coming from spent nuclear fuel. The paper describes studies for the latter MOX core model. The benchmark results include core criticality at the beginning and end of the equilibrium fuel cycle, kinetics parameters, spatial distributions of power, and reactivity coefficients obtained by employing different computation tools and nuclear data. Sensitivity studies were performed to better understand in particular the influence of variations in different nuclear data libraries on the computed results. Transient simulations were done to investigate the consequences of employing a few different sets of power and reactivity coefficient distributions on the system behavior. The obtained results are analyzed in the paper.     

Build-Up and Decay of Actinide Nuclides in Fuel Cycle of Nuclear Reactors

In order to assess the feasibility of utilizing plutonium in thermal reactors, build-up and decay of actinide nuclides have been studied for BWR, PWR, HWR, HTGR and LMFBR, which are uranium-oxide fueled or mixed-oxide fueled, and which produce electric power of 1,000MW. The following items were examined;

1. quantities of actinide nuclides build-up in the reactor

2. build-up and decay of activities of actinides in the spent fuel

3. build-up and decay of activities of actinides after reprocessing, and

4. variation of isotopie composition of plutonium with high burn-up.

It is concluded from the calculated results that precautions should be taken against high activities of resultant actinides if plutonium is utilized as a fissile material for thermal reactors. To make reprocessing and high-level waste management easy and practical, it is recommended that a thermal reactor should be fueled with uranium, the plutonium produced in a thermal reactor should be used in a fast reactor, and plutonium produced in the blanket of a fast reactor is more appropriate as fast reactor fuel than that from a thermal reactor.     

Scenario with inert matrix fuel: a specific study

In this paper, we describe a strategy study concerning the future of the French nuclear energy infrastructure, with a scenario involving reactors loaded with inert matrix fuel. We select the problem of the inventory control of minor actinides by target introduction into fast reactors. Added to pressurized water reactors in the French nuclear infrastructure, this scenario permits one to balance plutonium and minor actinide production and consumption and to obtain a substantial reduction of the radiological impact compared to a non-reprocessing fuel scenario on a one million year scale.     

Development of an automated system for fuel reload patterns design

In this paper an automated system to generate fuel reload patterns for a boiling water reactor (BWR), based on heuristic search methods using engineers expertise is presented. The main components of the system are the knowledge base, the inference engine, the 3D BWR simulator PRESTO-B and the user interface. The knowledge base includes a generation knowledge base and a modification knowledge base, which are concerned with the way human experts generate reload patterns. The system has been developed and applied to the Laguna Verde Nuclear Power Plant, achieving similar patterns to those used in the operation. No optimization algorithm has been incorporated in this system, therefore the generated reload patterns are the best estimate according the knowledge and experience of the nuclear engineers. Future works are being developed in this area using evolutionary optimization techniques as a complement of this system.     

Core reload optimization for equilibrium cycles using simulated annealing and successive linear programming

An algorithm is developed to determine directly all the parameters of the optimal equilibrium cycle. The core reload scheme is described by discrete variables, while the cycle length, as well as uranium enrichment and loading of burnable poison in each feed fuel assembly, are treated as continuous variables. An important feature of the algorithm is that all these parameters are determined by the solution of one big optimization problem. To search for the best reload scheme, simulated annealing is applied. The optimum cycle length as well as uranium enrichment and loading of burnable poison in each feed fuel assembly are determined for each reload pattern examined using successive linear programming. The uranium enrichments and loadings of burnable poison are considered to be distinct in different feed fuel assemblies. The number of batches and their sizes are not fixed, and also determined by the algorithm. As the first step of the numerical investigation of the algorithm, a problem of feed fuel cost minimization for a target equilibrium cycle length and fixed batch sizes is considered. The algorithm developed is demonstrated to provide about 2% less feed fuel cost than the ordinary simulated annealing algorithm.     

HTGR reactor physics and fuel cycle studies

The high-temperature gas-cooled reactor (HTGR) appears as a good candidate for the next generation of nuclear power plants. In the “HTR-N” project of the European Union Fifth Framework Program, analyses have been performed on a number of conceptual HTGR designs, derived from reference pebble-bed and hexagonal block-type HTGR types. It is shown that several HTGR concepts are quite promising as systems for the incineration of plutonium and possibly minor actinides.These studies were mainly concerned with the investigation and intercomparison of the plutonium and actinide burning capabilities of a number of HTGR concepts and associated fuel cycles, with emphasis on the use of civil plutonium from spent LWR uranium fuel (first generation Pu) and from spent LWR MOX fuel (second generation Pu). Besides, the “HTR-N” project also included activities concerning the validation of computational tools and the qualification of models. Indeed, it is essential that validated analytical tools are available in the European nuclear community to perform conceptual design studies, industrial calculations (reload calculations and the associated core follow), safety analyses for licensing, etc., for new fuel cycles aiming at plutonium and minor actinide (MA) incineration/transmutation without multi-reprocessing of the discharged fuel.These validation and qualification activities have been centred round the two HTGR systems currently in operation, viz. the HTR-10 and the HTTR. The re-calculation of the HTTR first criticality with a Monte Carlo neutron transport code now yields acceptable correspondence with experimental data. Also calculations by 3D diffusion theory codes yield acceptable results. Special attention, however, has to be given to the modelling of neutron streaming effects. For the HTR-10 the analyses focused on first criticality, temperature coefficients and control rod worth. Also in these studies a good correspondence between calculation and experiment is observed for the 3D diffusion theory codes.     

Performance improvement analysis of boiling water reactors by incorporation of hydride fuel

The feasibility of improving the neutronic characteristics of boiling water reactors (BWR) by using U–Zr hydride fuel is studied. Several modified BWR fuel assembly designs are considered. These include designs in which hydride fuel rods replace water rods only, replace water rods and a fraction of the oxide fuel rods, replace oxide fuel in the upper half of all the fuel rods, and replace all the oxide fuel in the assembly. It is found that replacement of at least half of the oxide fuel rods in the fuel assembly by U–ZrH_1.6 fuel might simultaneously improve the performance of BWR in three ways: (a) Increasing the energy extracted per fuel assembly and the cycle length by up to 10%. (b) Reducing the uranium ore and SWU requirements by approximately 10%. (c) Reducing the negative void coefficient of reactivity by, at least, 50%. It is also found that replacement of all the oxide fuel by hydride fuel opens interesting new options for the design of BWR fuel assemblies. The net result might be simplified assembly designs that can generate significantly more energy while featuring small negative void coefficient of reactivity. U–ThH₂ fuel appears to be even more promising than U–ZrH_1.6. For the potential benefits from hydride fuel to be realized, a clad material that is not permeable to hydrogen and is not as neutron absorbing as stainless steel needs to be developed.     

改进双排棒组件超临界水堆堆芯设计

为了提升堆芯性能,本文对现有的双排棒组件设计及堆芯设计方案进行了优化,并利用超临界核热耦合计算平台评估了优化后的方案。在组件设计中,为了减少寿期末堆芯中可燃毒物残余,优化了组件中可燃毒物棒的位置及可燃毒物含量。在堆芯设计中,为了延长堆芯寿期、降低包壳温度,对堆芯给水分配方案、换料方案及控制棒方案进行了一系列的优化。耦合计算结果表明,改进后的堆芯设计方案满足设计准则,堆芯寿期、卸料燃耗和包壳温度等参数均优于原方案。     

Criticality investigations for the fixed bed nuclear reactor using thorium fuel mixed with plutonium or minor actinides

Prospective fuels for a new reactor type, the so called fixed bed nuclear reactor (FBNR) are investigated with respect to reactor criticality. These are ① low enriched uranium (LEU); ② weapon grade plutonium + ThO₂; ③ reactor grade plutonium + ThO₂; and ④ minor actinides in the spent fuel of light water reactors (LWRs) + ThO₂. Reactor grade plutonium and minor actinides are considered as highly radio-active and radio-toxic nuclear waste products so that one can expect that they will have negative fuel costs.The criticality calculations are conducted with SCALE5.1 using S₈–P₃ approximation in 238 neutron energy groups with 90 groups in thermal energy region. The study has shown that the reactor criticality has lower values with uranium fuel and increases passing to minor actinides, reactor grade plutonium and weapon grade plutonium.Using LEU, an enrichment grade of 9% has resulted with k_eff = 1.2744. Mixed fuel with weapon grade plutonium made of 20% PuO₂ + 80% ThO₂ yields k_eff = 1.2864. Whereas a mixed fuel with reactor grade plutonium made of 35% PuO₂ + 65% ThO₂ brings it to k_eff = 1.267. Even the very hazardous nuclear waste of LWRs, namely minor actinides turn out to be high quality nuclear fuel due to the excellent neutron economy of FBNR. A relatively high reactor criticality of k_eff = 1.2673 is achieved by 50% MAO₂ + 50% ThO₂.The hazardous actinide nuclear waste products can be transmuted and utilized as fuel in situ. A further output of the study is the possibility of using thorium as breeding material in combination with these new alternative fuels.     

Effect of minor actinides doping on plutonium produced in large-scale FBR blanket

The present study focuses on the effect of minor actinides (MAs) addition into the FBR blanket as ways of increasing fraction of even-mass-number plutonium isotopes, especially ²³⁸Pu, aiming at enhancing the proliferation resistance of plutonium produced in the blanket. The MA loading potential to enhance the proliferation resistance of plutonium is investigated, with considering actual design constraints on the fuel decay heat from the fuel handling and fabrication points of view, as MAs considerably generate decay heat. It reveals that depending on doping quantity of MAs, it is possible to denature produced plutonium by MA transmutation. MA addition in the blanket gives a significant increment in ²³⁸Pu fraction of generated plutonium but less effect on other even-mass-number plutonium isotopes. However, it is important that MA compositions should be adequately controlled to satisfy both the proliferation resistance requirements and the decay heat constraints for fuel handling.