MOX Core Research Articles

Neutronic calculations for the VVER-1000 MOX core computational benchmark using the OpenMC code

The goal of this study is to perform neutronic calculations of the VVER-1000 MOX core computational benchmarks with an OpenMC code along with ENDF/B-VII.1 nuclear data library. The results of neutronic analysis using the OpenMC Monte Carlo code for the VVER-1000 MOX core, containing 30% mixed oxide fuel with low enriched uranium fuel, are presented in this study. As per the benchmark report, all six states are considered in the present study. The keff values, assembly average fission reaction rates, and pin-by-pin fission rates were calculated as per benchmark criteria. In addition, 2D thermal and fast neutron-flux distribution were also generated. The reactivity results and neutron flux distribution were compared with other results in which benchmark analysis was performed using the same core geometry and it showed great similarity with slight deviation. This shows that the modeling of the VVER-1000 MOX core was done successfully using OpenMC. Because OpenMC was successfully used for neutronics calculation of the VVER-1000 whole core, it may be mentioned here that OpenMC code can also be utilized for neutronics and other reactor core physics analyses of the VVER-1200 reactor which is to be commissioned in Bangladesh in the upcoming year.

A reaction rate measurement experimental program of China experimental fast reactor

The study of the nuclear reaction rate distribution in the MOX core of China Experimental Fast Reactor is a necessary condition for obtaining the operating license. From the characteristics of core structure, both the principle of experimental and experimental system are studies, emphasized on the scheme design such as the size of foils, the irradiation power and counting rate. Firstly, based on the Monte Carlo Code（MCNP）and NJOY codes, the distribution of reaction rates is obtained. MCNP is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron, and so on. The NJOY nuclear data processing system is a comprehensive computer code package for producing to cross sections and related nuclear parameters from ENDF/B VIII.0 evaluated nuclear data. Secondly, the activity measurement of foils was obtained by HPGE, then the reaction rate is determined by data processing. It is worth mentioning that the experimental reactivity introduction has a negligible effect on the reactor core by MCNP. Finally, a set of activation experiment scheme that is suitable for CEFR MOX core is formed, which serves as a guide to carry out activation method experiments.

Sub-channel analysis code development and core thermal calculation for fast reactor refueling cycle

Sub-channel analysis is an indispensable method for core safety evaluation. As a part of the fast reactor program series, development of SAC-SUB (System analysis code – SUB channel) program has been completed. Experiments performed with a 19-rod test assembly in the fuel failure mockup (FFM) sodium loop in Oak Ridge National Laboratory (ORNL) was selected as the benchmark to verify the SAC-SUB program capability. The maximum outlet temperature deviation is within 5%, which preliminarily verifies the availability of the program for fast reactor. Then the code was used to analyze CEFR thermal response during refueling cycle and MOX equilibrium state. During each refueling cycle, highest sodium temperature occurs in the fourth flow zone, and highest fuel temperature rise occurs in the first flow zone. The highest sodium temperature is 606.8 °C appearing in 3rd cycle, and the highest fuel temperature rise is 880.93 °C appearing in 4th cycle. Maximum temperature rise across the cladding and film do not change much. After reaching MOX core equilibrium state, all the key temperature will reach a lower level compared with refueling transition process, which indicating that the power distribution and flow distribution of the core are more matched in the equilibrium state.

Success-history based adaptive differential evolution method for optimizing fuel loading pattern of VVER-1000 reactor

A discrete Success-History based Adaptive Differential Evolution (SHADE) method has been developed and applied for fuel loading pattern (LP) optimization of a VVER-1000 reactor. SHADE uses an adaptive mechanism based on a historical record of successful control parameters, i.e., mutant scale F and crossover ratio CR, to improve the original differential evolution (DE) algorithm. Therefore, instead of three control parameters in the original DE, the SHADE method consists of two parameters of population size NP and memory size H. To apply the SHADE method to the fuel LP optimization, a relative position indexing approach was deployed to convert real variables into integer variables. Application of the discrete SHADE method to the problem of fuel LP optimization has been performed based on an VVER-1000 MOX core. A fitness function was chosen to maximize the keff, while flattening the radial power distribution. The results show that the keff of the optimal LP is greater than that of the reference core by about 1580 pcm. Whereas, the radial power peaking factor (PPF) of the optimal LP is about 2.4% smaller than that of the reference core.

Deterministic and Monte-Carlo interpretations of the MASURCA BALZAC-SI internal storage SFR experiment and quantification of uncertainties to nuclear data

Internal storage is used in fast reactors as it allows burnt fissile assemblies to cool down before being removed from the reactor. These assemblies are usually stored in the neutron shielding zone far from the core to avoid additional induced fissions. The BALZAC-SI experiment performed in the MASURCA Reactor was dedicated to study internal storage characteristics of a SFR-type MOX core. Fission rate distributions throughout the geometry are calculated with both the TRIPOLI-4® Monte-Carlo and the new APOLLO3® deterministic code. Results show that discrepancies up to 20% are obtained on fission rates with the TRIPOLI-4® code, the APOLLO3® code giving similar results to TRIPOLI-4®. Sensitivity and uncertainty calculations of the different responses to nuclear data are performed in order to explain these biases. We show that uncertainties due to nuclear data may explain a large part of the obtained biases, but further investigations are required to explain remaining discrepancies.

Evolutionary simulated annealing for fuel loading optimization of VVER-1000 reactor

An evolutionary simulated annealing (ESA) method has been developed for the problem of fuel loading optimization of VVER-1000 reactor. The ESA method improves original simulated annealing by using crossover and mutation operators to generate new trial loading patterns (LPs). A core physics calculation code for fuel LP optimization of VVER reactors (LPO-V) has been developed and verified based on a VVER-1000 MOX benchmark core in comparison with MCNP4c calculations. Calculations for optimizing fuel LP of the VVER-1000 MOX core have been conducted using the ESA method in comparison with simulated annealing (SA) and adaptive simulated annealing (ASA). Statistical differences between these methods were also evaluated based on the Mann–Whitney U test. The results show that the ESA method is advantageous over the SA and ASA.

Analysis of the methods for group constants generation for calculation of a large SFR core using Serpent 2 and CriMR codes

This work aimed at generating homogenized group constants using the Serpent code and then using the CriMR diffusion code to model the large SFR OECD 3600 MWth MOX core. The results were compared with a full core reference Monte Carlo solution by Serpent. Reactivity feedback parameters were also considered. Generating the group constants from separate fuel assemblies allows for simultaneously carrying out calculations and then using the results as input in diffusion codes rather than waiting so long for a 3D full core Monte Carlo calculation to be completed. From the results of the integral parameters we see a close agreement in the calculation codes. The differences can be attributed to the errors that could arise from generating the constants from individual sub-assemblies. The differences in the underlying physics and approximations used in development of the codes could also be a factor. Another way the errors could be reduced is by checking to see that the sub-assembly configurations used in the non-multiplying zones are as close as possible to the real layout in a full 3D core.

Analysis of VIP-BWR reactor core physics experiments on UO2 and MOX mockup fuel assemblies - part 2: analysis with SIMULATE5

ABSTRACT Analysis of the three test cores in the VIP-BWR program was performed in a two-dimensional geometrical model with CASMO5 coupled with the JENDL-4.0-based neutron data library, and reported in the previous paper. Following the study, interpretation of the experiments were carried out in a three-dimensional geometrical model with SIMULATE5 for the code validation study. The nuclear libraries for the SIMULATE5 calculations were generated with CASMO5 with the JENDL-4.0-based neutron data library. The effective multiplication factors of the critical cores ranged from 0.9983 to 1.0023 with measurement uncertainties of 0.0003 to 0.0004 (one σ). The root mean squares of (the calculated/the measured-1) for the fission rates at the core-mid plain of all the measured fuel rods were about 3% for the three cores. It was noticed that the calculations underestimated the fission rates of the UO2 fuel rods and overestimated those of the MOX fuel rods for the test cores loaded with MOX fuel rods, which was consistent with trends in the preceding analysis studies of the VIP-BWR program and other MOX core experiments, and the biases were confirmed in the calculation results of power distributions in MOX-fueled light water reactor cores.

Study on MOX Core Characteristics of Experimental Power Reactor using MCNP6 Code

In experimental power reactor (reaktor daya eksperimental, RDE), a full MOX fueled core can be accommodated in its operation without significant modifications on the reactor core design, but the MOX core tends to produce less favorable safety features and deliver transient behavior into unwanted fatal accidents. This paper aimed to investigate the MOX core characteristics of RDE through a series of calculations with MCNP6 code and ENDF/B-VII library. The calculation results show that MOX core with 235U enrichment of above 5 % can reach the criticality condition, maintain and run the reactor during the operation cycle. Utilizing MOX fuel in RDE with lower 235U enrichment will have a more negative impact on temperature coefficient of reactivity. The lower βeff makes the MOX core is more difficult to control, especially with low 235U enrichment. These results conclude that the selection of 235U enrichment in the MOX core must be carefully considered because it is one of the strategies to ensure the reactor safety criteria are fulfilled.

Generation of MCNP Whole-Core Reference Solutions for Advanced BWR MOX Core and Its Application to Neutronics Calculation Performance V&V of the BWR Nuclear Design Code

Generation of MCNP Whole-Core Reference Solutions for Advanced BWR MOX Core and Its Application to Neutronics Calculation Performance V&V of the BWR Nuclear Design Code

Effect of MOX fuel and the ENDF/B-VIII on the AP1000 neutronic parameters calculations by using MCNP6

The present work studies the effect of introducing MOX fuel on Westinghouse AP1000 neutronic parameters. The neutronic calculations were performed by using the MCNP6 code with the ENDF/B-VII.1 library and the new release of the ENDF/B-VIII, the AP1000 core with three 235U enrichment zones (2.35 %, 3.40 %, and 4.45 %). The obtained results showed that the simulated model for the AP1000 core satisfies the optimization criteria as a Westing- house reference. The results which included: effective multiplication factor, keff, delayed neutron fraction, beff, excess reactivity, rex, shutdown margin, temperature reactivity coefficients, whole core depletion, neutron flux, power peaking factor and core power density, were calculated and compared with the available published results. The keff in the cold zero power was found to be 1.20495 and 1.20247 with the ENDF/B-VII.1 and the ENDF/B-VIII libraries, respectively, which matches the value of 1.205 presented in the AP1000 Design Control Document for the UO2 fuel core. On the other hand, keff in the cold zero power was found to be 1.19988 and 1.19860 for MOX core with the ENDF/B-VII.1 and the ENDF/B-VIII libraries, respectively, which show good reception and confirm the safety of the design and efficient modeling of AP1000 reactor core.

Fuel loading pattern with 50% MOX fuel in HPR1000 core

Fuel loading pattern optimization with MOX (mixed oxide of plutonium and uranium) fuel is an important aspect for GIII LWR. For the HPR1000 core, an optimized MOX fuel is designed. Based on which, two loading patterns with 50% MOX fuel are designed. One is 1/4 core annually reloading pattern, and the other is 18-month reloading pattern. The core neutronic parameters are investigated by comparing with the 100% UO2 core. The MOX fuel has impacts on the nuclear characteristic parameters, but the 50% MOX core still satisfies the nuclear and safety criteria samilar to that considered in 100% UO2 core. The results of this study prove that the GIII LWR HPR1000 has excellent capability while loading with 50% MOX fuel.

Development of whole core pin-by-pin transport theory model in hexagonal geometry

The advances in computer processing power have made it possible to perform a detailed pin-by-pin calculation of the whole reactor core. A transport theory code TRANPIN has been developed to perform a whole reactor core pin-by-pin calculation in 2D hexagonal geometry. This code employs the interface current method based on 2D collision probability for reactor core calculation. The present code divides each lattice cell location in the fuel assembly (FA) into finer regions. The subdivided regions inside the lattice cell are connected using the 2D collision probabilities. The coupling of lattice cells within the assembly and assembly to assembly coupling is achieved using interface currents. The interface currents are obtained by expanding the angular flux leaving or entering the lattice cell surface into double orthonormal P2 polynomials. The interface current method is traditionally used to perform lattice calculations and its application to core calculations in hexagonal geometry is not reported in literature to the best knowledge of authors. The present code can perform the calculation in any multigroup energy structure. The code TRANPIN has been validated against two benchmark problems i) a simplified high temperature test reactor (HTTR) benchmark problem and ii) OECD VVER-1000 MOX Core Computational Benchmark. It may be mentioned that the OECD benchmark is analyzed using ultra fine WIMS library in 172 energy groups. This approach may be considered as unprecedented since the full reactor core calculations are normally performed with cross sections prescribed in few energy groups by some prior transport simulations. The present paper describes the interface current methodology incorporated in TRANPIN. The comparison of results of benchmark analyses with published results is also presented.

Representative neutronic characteristics calculations for the VVER-1000 reactors using SRAC and MCNP5

This paper presents the results of neutronic calculations using the deterministic and Monte-Carlo methods (the SRAC and MCNP5codes) for the VVER MOX Core Computational Benchmark Specification and the VVER-1000/V392 reactor core. The power distribution and keff value have been calculated for a benchmark problem of VVER core. The results show a good agreement between the SRAC and MCNP5 calculations. Then, neutronic characteristics of VVER-1000/V392 such as power distribution, infinite multiplication factor (k-inf) of the fuel assemblies, effective multiplication factor keff, peaking factor and Doppler coefficient were calculated using the two codes.

Impact of the thermal scattering law of H in H2O on the isothermal temperature reactivity coefficients for UOX and MOX fuel lattices in cold operating conditions

The contribution of the thermal scattering law of hydrogen in light water to isothermal temperature reactivity coefficients for UOX and MOX lattices was studied in the frame of the MISTRAL critical experiments carried out in the zero power reactor EOLE of CEA Cadarache (France). The interpretation of the core residual reactivity measured between 6 °C to 80 °C (by step of 5 °C) was performed with the Monte-Carlo code TRIPOLI4® . The nuclear data from the JEFF-3.1.1 library were used in the calculations. Three different thermal scattering laws of hydrogen in light water were tested in order to evaluate their impact on the MISTRAL calculations. The thermal scattering laws of interest were firstly those recommended in JEFF-3.1.1 and ENDF/B-VII.1 and also that recently produced at the atomic center of Bariloche (CAB, Argentina) with molecular dynamic simulations. The present work indicates that the calculation-to-experimpental bias is −0.4 ± 0.3 pcm/°C in the UOX core and −1.0 ± 0.3 pcm/°C in the MOX cores, when the JEFF-3.1.1 library is used. An improvement is observed over the whole temperature range with the CAB model. The calculation-to-experimpental bias vanishes for the UOX core (−0.02 pcm/°C) and becomes close to −0.7 pcm/°C for the MOX cores. The magnitude of these bias have to be connected to the typical value of the temperature reactivity coefficient that ranges from −5 pcm/°C at Begining Of Cycle (BOC) up to −50 pcm/°C at End Of Cycle (EOC), in PWR conditions.

Feedback on neutron capture cross sections of 238Pu and 241Am from analyses of measured isotopic compositions of irradiated LWR fuels and MOX core physics experiments

The atomic fractions of 238Pu and 241Am in MOX fuels recycled in light water reactors are 1% to 2% and not significant compared with those of major Pu isotopes. On the other hand, recent evaluated nuclear data libraries, such as JENDL-4.0 and JEFF-3.2, give noticeably different thermal and epithermal neutron capture cross sections for 238Pu and 241Am. The thermal neutron capture cross sections of 238Pu and 241Am in JEFF-3.2 are 31% and 9% larger than those of JENDL-4.0, respectively. This paper shows the effect of the differences in the neutron cross sections on analysis results of two different integral experiments. The first is the isotopic compositions of 238Pu on UO2 and MOX fuels irradiated in BWR and PWR, and the second is the critical experiments of the water moderated cores fully loaded with MOX fuels. The former was analyzed by using the continuous energy Monte Carlo burnup calculation code MVP-BURN and the latter by the continuous energy Monte Carlo calculation code MVP. The comparisons between the calculated and measured results indicate that the most likely thermal and epithermal neutron capture cross sections of 238Pu and 241Am should be around at the middle between those of JEFF-3.2 and JENDL-4.0.

Application of CNUREAS and MCNP5 codes to VVER-1000 MOX Core Computational Benchmark

In order to strengthen the nuclear design calculation capacity in Turkey, CNUREAS (Cekmece Nuclear Reactor System) was developed to provide easy usage of neutronic and thermal hydraulic nuclear codes included in the CNUREAS package. It was tested and used for research reactors and PWR type power reactors. Modifications were performed to add hexagonal geometry support taking into account VVER type reactors employing hexagonal fuel assemblies that will be built in Turkey. “VVER-1000 MOX Core Computational Benchmark” was used to test new features of the CNUREAS. The maximum deviation in effective multiplication factor results of CNUREAS was 0.7% with deterministic codes and 1.5% with Monte Carlo codes. It was concluded that CNUREAS can be used for neutronic calculations of VVER type power reactors with appropriate cross section libraries and deterministic and Monte Carlo techniques give comparable results when both provided with appropriate cross section libraries.

Validation of 241Am Capture Cross Section through Integral Test Using Criticality Data of Light-water Moderated MOX Cores

Abstract In the development of JENDL-4.0, the thermal and resonance capture cross section of 241 Am was increased by about 8-9% to reduce the overestimation of criticality observed in the MOX-Fueled light-water-moderated cores using 241 Am-rich plutonium-fuel.The revised data gives the better results of criticality prediction for those cores than that with JENDL-3.3. The Pu ageing problem was also investigated using the revised data. This problem is that the calculated k-effs for critical MOX cores with different experiment dates showed the tendency of increase with time. The change of k-eff in seven years was about 0.2%dk/k with JENDL-3.3 and it was reduced to about 0.1%dk/k with JENDL-4.0. These results show that the revision of 241 Am capture cross section seemed to be successful, although there still remains the tendency of k-eff increase with time. On the other hand, the recent measurement of thermal and resonance capture cross section of 241 Am indicates that the data of JENDL-3.3 should be increased by about 18%, that is about 8% greater than that of JENDL-4.0. In the present study, the possibility of further revision of 241 Am capture cross section was studied through the integral test for Pu ageing problem by reflecting the recent experimental data. Three criticality data of a light-water moderated MOX core with different experiment date (7 years interval at maximum) have been analyzed with JENDL-3.3 and 4.0 libraries. In addition, the calculations with the number density of 241 Am multiplied by 1.18 for JENDL-3.3 and by 1.08 for JENDL-4.0 were also conducted to simulate the increase of 241 Am capture cross section suggested by the recent measurement. The results show that the increase tendency of k-eff as a function of time observed in the calculations with JENDL-4.0 is not found in those with modified 241 Am number density. This means that the improvement of 241 Am capture cross section will be achieved by employing the recent experimental data.

Challenge of PWR new core design simulation: A focus on uncertainties due to nuclear data and reflector modelling

This paper presents a reactor core uncertainty analysis in the framework of the OECD/NEA UAM Benchmark. Three types of uncertainties affecting the predictions of power distribution in the core of a nuclear reactor are discussed: the uncertainties of basic nuclear data, the uncertainties resulting from the use of different simulation tools and those due to approximations in reflector modelling. The contribution of nuclear data uncertainty on the power distribution of a UOX and a MOX core is assessed with the XSUSA tool. Overall, the results obtained with different tools in both institutions are in good agreement, showing that the power distribution uncertainty due to the use of different simulation tools is much lower than the one due to nuclear data, which is a large contributor. Lastly, the paper presents preliminary work showing the relevance of reflector modelling on the uncertainty of the power distribution at nominal conditions as well as on an asymmetrical case representative of accidental conditions.

Robust, Highly Thermally Stable, Core–Shell Nanostructured Metal Oxide Aerogels as High-Temperature Thermal Superinsulators, Adsorbents, and Catalysts

Robust, highly thermally stable, MOx/(MOx–SiO2)/SiO2 core–shell nanostructured metal oxide aerogels with a MOx core and (MOx–SiO2)/SiO2 shell are produced via novel alkoxide chemical liquid deposition techniques. The core–shell nanostructure not only significantly reinforces the nanoparticles but also effectively inhibits the crystal growth and phase transition of metal oxide upon heat treatment, which enhances the heat resistance from approximate 400–800 °C up to 1000–1300 °C. The resultant core–shell nanostructured Al2O3, ZrO2, and TiO2 aerogels can support at least 5800 times their weight and exhibit high surface areas of 139, 186, and 154 m2/g after fired at 1300, 1000, and 1000 °C, respectively, which are the highest surface areas for metal oxide aerogels ever reported. We demonstrate that the core–shell ZrO2 and TiO2 aerogels show enhanced adsorption and photocatalytic performances, respectively, for dye after fired at 1000 °C. The core–shell Al2O3 aerogel/mullite fiber/TiO2 composite possesses ultralow thermal conductivities of 0.058, 0.080, and 0.11 W/mK at 800, 1000, and 1200 °C, respectively, which are the lowest values for inorganic aerogels ever reported. The resulting materials are promising candidates as high-temperature (400–1300 °C) thermal superinsulators, adsorbents, and catalysts.