MOX Core Research Articles

VVER-1000 MOX Core Computational Benchmark

This report presents the VVER MOX Core Computational Benchmark Specification and Results, which was proposed as a benchmark within the OECD/NEA Expert Group on Reactor-based Plutonium Disposition (TFRPD). Benchmark results, obtained using three computer codes, are presented. The codes include: the MCU Monte Carlo code (Kurchatov Institute, Russian Federation), RADAR (Kurchatov Institute, Russian Federation), and the MCNP Monte Carlo code (GRS, Germany). The codes use different methods and different nuclear data. A comparison of the results shows good agreement among the various codes.

A computationally simple model for determining the time dependent spectral neutron flux in a nuclear reactor core

The balance of isotopes in a nuclear reactor core is key to understanding the overall performance of a given fuel cycle. This balance is in turn most strongly affected by the time and energy-dependent neutron flux. While many large and involved computer packages exist for determining this spectrum, a simplified approach amenable to rapid computation is missing from the literature. We present such a model, which accepts as inputs the fuel element/moderator geometry and composition, reactor geometry, fuel residence time and target burnup and we compare it to OECD/NEA benchmarks for homogeneous MOX and UOX LWR cores. Collision probability approximations to the neutron transport equation are used to decouple the spatial and energy variables. The lethargy dependent neutron flux, governed by coupled integral equations for the fuel and moderator/coolant regions is treated by multigroup thermalization methods, and the transport of neutrons through space is modeled by fuel to moderator transport and escape probabilities. Reactivity control is achieved through use of a burnable poison or adjustable control medium. The model calculates the buildup of 24 actinides, as well as fission products, along with the lethargy dependent neutron flux and the results of several simulations are compared with benchmarked standards.

Advanced MOX Core Design Study of Sodium-Cooled Reactors in Current Feasibility Study on Commercialized Fast Reactor Cycle Systems in Japan

Sodium-cooled mixed-oxide core design studies are performed with a target burnup of 150 GWd/t and possible measures against the recriticality issues in core disruptive accidents. Four types of core are compared from the viewpoints of core performance and reliability. Results show that all the types of core satisfy the target and that a homogeneous core with an axial blanket partial elimination subassembly is the superior concept, although experimental demonstration is required of molten fuel motion for mitigation of recriticality following fuel melting and loss of fuel pin integrity.

380. MOX燃料炉物理試験MISTRALの解析

Nuclear Power Engineering Corporation (NUPEC) has been performing conceptual design studies of high moderation full MOX LWR cores that aim for increasing fissile Pu consumption rate and reducing residual Pu in discharged MOX fuel. As part of these studies, NUPEC, French Atomic Energy Commission (CEA) and their industrial partners implemented an experimental program, MISTRAL, that was devoted to measuring the core physics parameters of such advanced cores. The program consists of one reference UO2 core, two homogeneous full MOX cores and one full MOX PWR mock-up core that have higher moderation ratio than the conventional lattice.NUPEC has been analyzing the experimental results with the diffusion and the transport calculations by the SRAC code system and the continuous energy Monte Carlo calculations by the MVP code with the common nuclear data file, JENDL-3.2. The calculation results well reproduce the experimental data approximately within the same range of the experimental uncertainty. This indicates that these applied analysis methods give the same accuracy for the UO2 core and MOX cores, for the different moderation MOX cores, and for the homogeneous and the mock-up MOX cores.

Analysis of MISTRAL Experiments with JENDL-3.2

Nuclear Power Engineering Corporation has been analyzing measurement data of MISTRAL, a LWR MOX core physics experimental program, with a deterministic code system of the diffusion and the transport theories, SRAC, and a Monte Carlo transport code, MVP, based on JENDL-3.2. The program consists of one reference U02 core, two homogeneous full MOX cores and one full MOX PWR mock-up core that have higher moderation ratio than the conventional lattice. The measurement parameters cover critical masses, boron concentrations, radial and axial core power distribution, spectrum indexes, conversion factors, effective delayed neutron fractions, iso-thermal temperature coefficients, boron efficiency, single absorber worth, water hole worth, absorber cluster worth and void worth. Critical keff calculations show a slight overestimate for MOX cores and a small systematic trend with aging of MOX fuel. Power distribution calculations show good agreement with the measurements in both homogeneous and mock-up cores. Measured spectrum indexes of fissile isotopes are well reproduced by the calculation. For other measurements, calculated results well reproduce the measurements within one or two times of the experimental error. The analysis verifies the validity of JENDL-3.2 and the analysis methods for the high moderation LWR MOX cores.

Analysis of High Moderation PWR MOX Core MISTRAL-4 with SRAC and MVP

An extensive experimental program, MISTRAL, is undertaken by NUPEC and CEA in order to measure the main core physics parameters of high moderation 100% MOX LWRs. The analysis of the MISTRAL-4, a mock-up of a full MOX PWR core was carried out by diffusion/transport calculations with the SRAC code system and by continuous-energy Monte Carlo calculations with the MVP code. The calculation results agreed with the experimental results of both calculations within experimental uncertainties and calculation results showed no specific trend caused by heterogeneity in highly-moderated mock-up core configurations.

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.

A numerical neutronics benchmark study for inert matrix plutonium fuels in uranium dioxide and mixed plutonium-uranium dioxide environments

A numerical benchmark exercise has been under way for comparing the results of different calculational methods/data sets used for the analysis of light water reactor (LWR) configurations employing Pu inert matrix fuels (IMFs). The first phase of the exercise was devoted to infinite arrays of identical IMF cells. The main feature investigated in the second phase has been the influence of the neutron spectra in UO 2 and MOX cores on individual IMF cells. Phase 3 is concerned with the more realistic situation of an IMF assembly surrounded by UO 2 assemblies. Significant discrepancies have been observed for power peaking effects and delayed neutron parameters in Phase 2. In Phase 3, neutron balance differences for the IMF, particularly at EOL, are found to be significantly larger than were observed in Phase 1.

研究論文 １１１３． 軽水炉でのプルトニウム有効利用に向けた高減速ＭＯＸ炉心

Conceptual design studies have been performed for high moderation full MOX cores aiming at increasing fissile Pu consumption rate (ratio of the consumed to the loaded fissile Pu) and reducing residual Pu in discharged MOX fuel. The BWR cores studied have hydrogen to heavy metal ratio (H/HM) of 5.9 with increasing water rods and 7.0 with reducing a fuel rod diameter based on a reference 9×9 fuel (H/HM=4.9) of ABWR. The PWR cores studied have H/HM of 5.0 and 6.0 with reducing a fuel rod diameter based on a reference 17×17 fuel (H/HM=4.0) of APWR. Equilibrium core design and plant safety analyses showed that those high moderation cores have compatibility with ABWR and APWR. The fissile Pu consumption rate is 22% larger than the full MOX cores with reference fuel of ABWR and 50% for APWR. The core performance and compatibility has been also evaluated in the condition of multi-recycle of Pu in these high moderation cores.Study has been conducted to evaluate the effect of introducing these high moderation cores in the fuel cycle of Japan. It shows that the high moderation cores produce 26% more cumulative electricity and reduce 22% stock of the fissile Pu by 2050 than the reference cores.

軽水流使用済燃料の長期貯蔵がプルサーマル燃料サイクルに与える効果

According to the Long-term Program for Research, Development and Utilization of Nuclear Energy (June, 1994) in Japan, the Rokkasho Reprocessing Plant will be operated shortly after the year 2000, and the planning of the construction of the second commercial plant will be decided around 2010. Also, it is described that spent fuel storage has a positive meaning as an energy resource for the future utilization of Pu.Considering the balance between the increase of spent fuels and the domestic reprocessing capacity in Japan, it can be expected that the long-term storage of UO2 spent fuels will be required.Then, we studied the effect of long-term storage of spent fuels on Pu-thermal fuel cycle. The burnup calculation were performed on the typical Japanese PWR fuel, and the burnup and criticality calculations were carried out on the Pu-thermal cores with MOX fuel. Based on the results, we evaluate the influence of extending the spent fuel storage term on the criticality safety, shielding design of the reprocessing plant and the core life time of the MOX core, etc.As the result of this work on long-term storage of LWR spent fuels, it becomes clear that there are few demerits regarding the lifetime of a MOX reactor core, and that there are many merits regarding the safety aspects of the fuel cycle facilities. Furthermore, long-term storage is meaningful as energy storage for effective utilization of Pu to be improved by technological innovation in future, and it will allow for sufficient time for the important policymaking of nuclear fuel cycle establishment in Japan.

Concept and basic performance of an in-pile experimental reactor for fast breeder reactors using fast driver core

The possibility of an in-pile experimental reactor for fast breeder reactors using a fast driver core is investigated. The driver core is composed of a particle bed with diluted fuel. The results of various basic analyses show that this reactor could perform as follows: (1) power peaking at the outer boundary of test core does not take place for large test core; (2) the radial power distribution in test fuel pin is expected to be the same as a real reactor; (3) the experiments with short half width pulse is possible; (4) for the ordinary MOX core, enough heating-up is possible for core damage experiments; (5) the positive reactivity effects after power burst can be seen directly. These are difficult for conventional thermal in-pile experimental reactors in large power excursion experiments. They are very attractive advantages in the in-pile experiments for fast breeder reactors.

96/00392 Plant safety — Pool boiling issue still not settled for engineer

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 βeff 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 βeff 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.