WIMSD-5B Codes Research Articles

Neutronic Analysis of the RSG-GAS Fuel Using Burnable Poison

Control and safety of nuclear reactors are significantly influenced by the determination of safety parameters. The three most crucial safety factors for assessing reactor status are the infinite multiplication factor, reactivity coefficients, and power peaking factor. The objective of the present study is to examine how the RSG-GAS fuel safety parameters behave in a typical reactor operation state. A lattice cell fuel model of the fuel lattice of the RSG-GAS reactor core was modeled using WIMSD-5Bwith cross-section library data based on ENDF/B-VIII.0. The value of the infinite multiplication factor with various burnable poison concentrations, as well as the moderator and fuel temperatures, were the variables that were examined. The reactivity coefficient parameters were similarly analyzed. By comparing the WIMSD-5B code results with information from the SAR document, the WIMS model for RSG-GAS fuel was verified, and it was inferred that the parameters are in good agreement. Safe behavior uses the predicted reactivity coefficient values as an example.

NEUTRONIC ANALYSIS OF THE VVER-1200 LATTICE CELL FUEL USING WIMSD-5B CODE

The calculation of safety parameters in nuclear reactors has an important influence on nuclear reactor control and safety. The infinite multiplication factor, reactivity coefficients, and power peaking factor parameters are the most important safety parameters for determining reactor status. The aim of the present study is to analyze the behavior of the nuclear safety parameters for the VVER-1200 core in a normal state of reactor operation. A lattice cell fuel model of the VVER-1200 reactor core was performed using WIMSD-5B. The cross-section library data based on the ENDF/B-VIII.0 was used. The investigated parameters were the value of infinite multiplication factor with different pitch, temperature, enrichment, and boron concentration. The calculation also investigated the reactivity coefficient parameters. The verification of WIMS model VVER-1200 was performed by comparing the results of the WIMSD-5B code with VVER-1200 data in the SAR document, and it was implied that they are in good agreement. The calculated values of reactivity coefficients illustrated a safe behavior.

ANALYSIS OF THORIUM PIN CELL BURN UP OF THE PWR USING WIMS CODE

A thorium-fueled benchmark comparison was made in this study between state-of-the-art codes, WIMSD-5B code to MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations as part of efforts to examine the possible benefits of using thorium in PWR fuel. WIMSD-5B calculations employ the same model as a reference, MOCUP, and CASMO, however, there are some variances in methodology and cross-section libraries. On a PWR pin cell model, eigenvalue and isotope concentrations were examined up to high burnup. The eigenvalue comparison as a function of burnup is good, with a maximum difference of less than 5% and an average absolute difference of less than 1%. The isotope concentration comparisons outperform a set of ThO2-UO2 fuel benchmarks and are comparable to a set of uranium fuel benchmarks previously published in the literature. As a function, the eigenvalue comparison The actinide and fission product data sources for a typical thorium fuel are reported in the WIMSD-5B burnup calculations. The reasons for discrepancies in coding are examined and explored.Keywords: Thorium, PWR Fuel, Burn up, Pin Cell, WIMSD-5B

NEUTRONIC CALCULATION FOR PWR MOX FUEL PIN CELLS WITH WIMSD-5B CODE

NEUTRONIC CALCULATION FOR PWR MOX FUEL PIN CELLS WITH WIMSD-5B CODE. The WIMSD-5B thermal reactor lattice cell code is used in many laboratories for research reactor calculations and power reactors. The program uses the Wigner-Seitz approximation for cell pin calculations. The approximation has been widely applied to the pin of UO2 cells and has shown good results in previous studies but can produce incorrect results when used for pin cells in MOX fuels. This paper investigates the use of the WIMS-5B code to calculate the neutron multiplication factor and depletion for MOX fuel pin cells. Calculations were performed using the WIMSD-5B code updated with the ENDF-BVIII.0 library. The outer scattering boundary condition was used to overcome the effect of the Wigner-Seitz approach on the lack of MOX fuel. Results of this study indicates that most of the results obtained using ENDF-BVIII.0 are better than ENDF-BVII.1, and this is in line with expectations. The difference in the maximum k-inf value obtained from this library occurs in the fuel that has the greatest enrichment. On the other hand, the addition of the outer scattering limit improves the results obtained using ENDF-BVIII.0, causing a slight improvement for other libraries. This shows that by using appropriate libraries and the addition of the scattering outer limit, the Wigner Seitz approximation for the MOX pin cell pins in WIMS-D5 can yield quite accurate results.Keywords: Wigner-Seitz approximation, WIMS-D5 code, MOX fuel, Doppler reactivity.

Simulation of the burnup in cell calculation using the wimsd-5b code considering different nuclear data libraries

This work proposes to implement the cell calculation considering the fuel burning using the WIMSD-5B code. The cell calculation procedure allows determining the nuclear parameters present in the multi-group neutron diffusion equation and for this purpose the neutron transport theory is used in a problem with dimensional reduction, but in contrast is considered a large number of groups associated with the neutron spectrum. There are a variety of reactor physics codes that determine the nuclear parameters by solving the neutron transport equation applied to an equivalent cell representing a fuel element. The WIMSD-5B code is a deterministic code that solves the transport equation using collision probability method. The simulation of fuel burning in the cell calculation took into account different nuclear data libraries. The WIMSD-5B code supports several nuclear data libraries and in the present work the following libraries were used: IAEA, ENDFB-VII.1, JENDL3.2, JEFF3.1 and JEF2.2, all formatted for 69 energy groups.

Calculation of PWR Thorium Pin Cell Burnup and Isotope Prediction Using WIMSD-5B Code

The amount of thorium in the world is very abundant and the benefits of using thorium as fuel for nuclear reactors such as PWR are increasingly widespread, so it needs to be analyzed. In the initial phase, the PWR pin cell fuel analysis was carried out using the WIMSD-5B code. The PWR ThO2-UO2 fuel in the cladding with the fuel assembly model 17 x 17 is analyzed by its fuel burn up, neutron flux and isotope composition after maximum operation. The pin cell model consisting of fuel, cladding and coolant with a square pitch of 12.6 cm are calculated using different libraries. Eigenvalues, neutron fluxes and isotope concentrations were compared with the PWR pin cell model to high burnup. The eigenvalues and flux neutron as a function of burnup are good, the maximum difference is within 7.41% and the mean absolute difference is less than 3.12 %. The best comparison of fission product isotope concentrations is to use the ENDFB7.1 library with Thorium fuel and is comparable to the uranium fuel reported in the literature. Data sources for actinides and fission products used infuel depletion calculations for thorium fuel are all documented.

Progress on China nuclear data processing code system

China is developing the nuclear data processing code Ruler, which can be used for producing multi-group cross sections and related quantities from evaluated nuclear data in the ENDF format [1]. The Ruler includes modules for reconstructing cross sections in all energy range, generating Doppler-broadened cross sections for given temperature, producing effective self-shielded cross sections in unresolved energy range, calculating scattering cross sections in thermal energy range, generating group cross sections and matrices, preparing WIMS-D format data files for the reactor physics code WIMS-D [2]. Programming language of the Ruler is Fortran-90. The Ruler is tested for 32-bit computers with Windows-XP and Linux operating systems. The verification of Ruler has been performed by comparison with calculation results obtained by the NJOY99 [3] processing code. The validation of Ruler has been performed by using WIMSD5B code.

Improving the refueling cycle of a WWER-1000 using cuckoo search method and thermal-neutronic coupling of PARCS v2.7, COBRA-EN and WIMSD-5B codes

The fuel loading pattern optimization is an important process in the refueling design of a nuclear reactor core. Also the analysis of reactor core performance during the operation cycle can be a significant step in the core loading pattern optimization (LPO). In this work, for the first time, a new method i.e. cuckoo search algorithm (CS) has been applied to the fuel loading pattern design of Bushehr WWER-1000 core. In this regard, two objectives have been chosen for finding the best configuration including the improvement of operation cycle length associated with flattening the radial power distribution of fuel assemblies. The core pattern optimization has been performed by coupling the CS algorithm to thermal-neutronic codes including PARCS v2.7, COBRA-EN and WIMSD-5B for earning desired parameters along the operation cycle. The calculations have been done for the beginning of cycle (BOC) to the end of cycle (EOC) states. According to numerical results, the longer operation cycle for the semi-optimized loading pattern has been achieved along with less power peaking factor (PPF) in comparison to the original core pattern of Bushehr WWER-1000. Gained results confirm the efficient and suitable performance of the developed program and also the introduced CS method in the LPO of a nuclear WWER type.

Reactivity Prediction Calculations for Infinite Arrays of PWR MOX Fuel Pin Cells using MCNP5 and WIMSD-5B

In the present paper, the effect of some major nuclides on the reactivity of nuclear fuel for fuel pin cell in the Pressurised Water Reactor (PWR) was calculated using MCNP5_RSICC, 1.2 and WIMSD-5B codes. In order to assess our capabilities, the credit criticality Benchmark for infinite arrays of PWR MOX Fuel Pin Cells were calculated using two different codes and two different cross section libraries. Results were compared with similar published results by the Organization for Economic Co-operation and Development (OECD/NEA) Expert Group on Burn-up Credit. Good agreement was found for the value of keff with the corresponding codes and the reference mean solutions. The reactivity worth due to uranium, plutonium and curium isotopes was determined. The discrepancies between the present results and the reference mean results are within the range of the discrepancies for the participants with the reference mean results. The spread in the keff results with the mean results varies from -0.246 to 0.276 and from 0.05 to 0.793 for MCNP5 and WIMSD-5B respectively.

A NOVEL APPROACH TO FIND OPTIMIZED NEUTRON ENERGY GROUP STRUCTURE IN MOX THERMAL LATTICES USING SWARM INTELLIGENCE

Energy group structure has a significant effect on the results of multigroup transport calculations. It is known that UO2–PUO2 (MOX) is a recently developed fuel which consumes recycled plutonium. For such fuel which contains various resonant nuclides, the selection of energy group structure is more crucial comparing to the UO2 fuels. In this paper, in order to improve the accuracy of the integral results in MOX thermal lattices calculated by WIMSD-5B code, a swarm intelligence method is employed to optimize the energy group structure of WIMS library. In this process, the NJOY code system is used to generate the 69 group cross sections of WIMS code for the specified energy structure. In addition, the multiplication factor and spectral indices are compared against the results of continuous energy MCNP-4C code for evaluating the energy group structure. Calculations performed in four different types of H2O moderated UO2–PuO2 (MOX) lattices show that the optimized energy structure obtains more accurate results in comparison with the WIMS original structure.

Calculation of the fuel composition and the thermo-neutronic parameters of the Bushehr’s VVER-1000 reactor during the initial startup and the first cycle using the WIMSD5-B, CITATION-LDI2 and WERL codes

In this paper, the concentrations of fission products and fuel isotopes as well as the changes of the thermo-neutronic parameters of the Bushehr’s VVER-1000 reactor were studied during the initial startup and the first cycle. In order to perform the time-dependent cell calculations and obtain the concentration of fuel elements, the WIMSD5-B code was used. Besides, by utilizing the CITATION-LDI2 code, the effective multiplication factor and the thermal power distribution of the reactor were calculated. A computer program (WERL code) was designed in order to perform accurate calculation of the temperature distribution of the reactor core. For this purpose, the Ross–Stoute, Weisman, and Lee–Kesler models were used for calculating of the gap conductance coefficient, fission gas release and gap pressure, respectively. The results demonstrated that in designing the startup process, in addition to the role considered for overcoming the power defects and in preparing the required conditions for performing the safety-assurance tests, the flattening of the reactor’s power must be taken into account. Comparison between the results of this modeling and the final safety analysis report of this reactor showed that the results presented in this paper are satisfactorily accurate.

Impact assessment of upscattering on resonance calculation using improved ultrafine energy group method

Thermal upscattering has considerable impact on the neutronic analysis of light water reactors. This phenomenon is more significant for reactors with advanced fuels having high plutonium content: Mixed Oxide (MOX) fuels. In this paper, a sensitivity analysis for the impact of upscattering boundary energy on the integral results of MOX and UO2 fuels in light water reactors is performed using WIMSD-5B code. For investigating the impact on the effective cross sections, a spectrum calculation code Microtrans is developed which employs an improved ultrafine group method. Moreover, the impact on the effective cross sections of important resonances of fuel nuclides in thermal and epithermal energy ranges is studied in detail. Calculation results show that changing the upsacttering boundary from 0.625eV to 4eV leads to more than 1000pcm decrease in Keff in MOX fuels due to large resonances of Pu isotopes in thermal energy range, while this impact is about 120pcm for UO2 fuels. In addition, thermal upscattering has significant impact on the effective cross sections below 10eV.

A methodology to a DB-MHR fuel recharge evaluation—A basic comparison between WIMSD-5B and MCNPX codes

The deep burn modular helium reactor (DB-MHR) belongs to the new generation of nuclear power plants. Such reactor presents the possibility to use reprocessed fuels and a special refueled based on shuffling scheme that displaces the fuel pins after each cycle. This work presents a simplified methodology to evaluate qualitatively the shuffling scheme of these reactors using the WIMSD-5B code. The keff behavior during the burnup was verified by the WIMSD-5B and the MCNPX 2.6.0 codes using three types of core geometry. The WIMD-D5 and the MCNP presented very similar results.

A comparative study of integral parameters for TRX and BAPL benchmark lattices of thermal reactors using the cross-section library generated from JENDL and JEF evaluated data files

The objective of this paper is to present the results of comparative study of integral parameters for TRX and BAPL benchmark lattices of thermal reactors. The nuclear data processing code NJOY'99 was deployed for the generation of the 69-group cross-section library from the basic evaluated nuclear data files JENDL-3.2 and JEF-2.2. TRX and BAPL benchmark lattices were modeled with optimized inputs, which were suggested in the final report of the WIMS Library Update Project Stage-I. The inputs were the results of a detailed parametric study of the WIMS input options and also optimized for accuracy. The integral parameters (such as k eff, ρ 28, δ 25, δ 28, C ∗) of five uranium-fuel thermal assemblies: TRX-1 and TRX-2 and BAPL-1, BAPL-2, and BAPL-3 were calculated with the help of WIMSD-5B code based on the generated 69-group cross-section library. The calculated results are compared with those of experiments and it is found that the obtained results between the two libraries are in good agreement with each other. Besides, the calculated integral parameters are also well consistent with the measured values, which reflect the validation of the generated 69-group cross-section library and this library thus obtained is necessary to meet up the nuclear data for neutronics calculation of TRIGA Mark-II research reactor at AERE, Savar, Dhaka, Bangladesh.

On collapsing the [formula omitted] average number of neutrons released per fission from the IAEA.LIB library with the WIMSD-5b code

It was verified after a fuel burnup calculation with the WIMSD-5b code using the IAEA.LIB library that the computed average number of neutrons released per fission of Pu 94 242 shows up as a Not-a-Number (NaN) for some energy groups. As this problem does not permit the use of the generated multigroup microscopic cross sections by a reactor calculation code, the value of 1.0E−38 barns was attributed to all energy groups of the IAEA.LIB library that have null values of multigroup microscopic fission cross sections for this material.