Computer Code System Research Articles

Development and assessment of ASTEC-Na fuel pin thermo-mechanical models performed in the European JASMIN project

ASTEC-Na is a computer code system which evaluates protected and unprotected accidents in Sodium-cooled Fast Reactors throughout the Initiation Phase. The fuel pin behavior models implemented in ASTEC-Na simulate the essential aspects of the thermal and mechanical behavior of SFR fuel pins in nominal and accidental conditions. Besides the improvement of existing ASTEC-Na models, a specific cladding mechanical model and an in-pin fuel relocation model have been developed during the JASMIN project, where the latter has been already implemented in the code. ASTEC-Na fuel pin behavior models are described in this paper and their transient predictions for four CABRI transient tests are compared to available experimental data as well as to SAS-SFR and SIMMER-III code calculations. Conclusions on the overall performance of ASTEC-Na fuel pin models are also presented.

Development of three-dimensional reactor analysis code system for accelerator-driven system, ADS3D and its application with subcriticality adjustment mechanism

ABSTRACTIn order to perform the parametric survey for an accelerator-driven system (ADS) core with the subcriticality adjustment mechanism, a new calculation code system, ADS3D, was developed on MARBLE which is a comprehensive and versatile framework for reactor analysis. The application of ADS3D was also demonstrated on the neutronics design of ADS operated by control rod (CR) movement. Through the neutronics calculation, it was shown that the maximum proton beam current was decreased from 20.5 to 11.6 mA due to the switch from beam-operated to CR-operated core.

Analysis of neutronics benchmarks for the utilization of mixed oxide fuel in light water reactor using DRAGON code

Advances in reactor physics have led to the development of new computational technologies and upgraded cross-section libraries so as to produce an accurate approximation to the true solution for the problem. Thus it is necessary to revisit the benchmark problems with the advanced computational code system and upgraded cross-section libraries to see how far they are in agreement with the earlier reported values. Present study is one such analysis with the DRAGON code employing advanced self shielding models like USS and 172 energy group ‘JEFF3.1’ cross-section library in DRAGLIB format. Although DRAGON code has already demonstrated its capability for heavy water moderator systems, it is now tested for light water reactor (LWR) and fast reactor systems. As a part of validation of DRAGON for LWR, a VVER computational benchmark titled “Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel-Volume 3” submitted by the Russian Federation has been taken up. Presently, pincell and assembly calculations are carried out considering variation in fuel temperature (both fresh and spent), moderator temperatures and boron content in the moderator. Various parameters such as infinite neutron multiplication (k∞) factor, one group integrated flux, few group homogenized cross-sections (absorption, nu-fission) and reaction rates (absorption, nu-fission) of individual isotopic nuclides are calculated for different reactor states. Comparisons of results are made with the reported Monte Carlo (MCU) values of the benchmark. Maximum deviation of 1.8% in k∞ is observed for variants with spent fuel and for the states with control rod whereas all the other results are in par with the results reported in the benchmark document. The few and multi-group macroscopic cross-sections and flux of all the nuclides also compare well with the benchmark results except for the 11B macroscopic absorption cross section, which is further compared with the XNWLUP software. Inter-comparison of results with the generalized self-shielding model SHI of DRAGON code employing the traditional WIMSD formatted 172 group cross-section library has also been made to highlight the improvements made in computational schemes and cross-section library format.

The CCONE Code System and its Application to Nuclear Data Evaluation for Fission and Other Reactions

A computer code system, CCONE, was developed for nuclear data evaluation within the JENDL project. The CCONE code system integrates various nuclear reaction models needed to describe nucleon, light charged nuclei up to alpha-particle and photon induced reactions. The code is written in the C++ programming language using an object-oriented technology.At first, it was applied to neutron-induced reaction data on actinides, which were compiled into JENDL Actinide File 2008 and JENDL-4.0. It has been extensively used in various nuclear data evaluations for both actinide and non-actinide nuclei. The CCONE code has been upgraded to nuclear data evaluation at higher incident energies for neutron-, proton-, and photon-induced reactions. It was also used for estimating β-delayed neutron emission.This paper describes the CCONE code system indicating the concept and design of coding and inputs. Details of the formulation for modelings of the direct, pre-equilibrium and compound reactions are presented. Applications to the nuclear data evaluations such as neutron-induced reactions on actinides and medium-heavy nuclei, high-energy nucleon-induced reactions, photonuclear reaction and β-delayed neutron emission are mentioned.

Modelling and analysis of nucleon emission from deuteron-induced reactions at incident energies up to 100 MeV

We have so far developed a computational code system dedicated to deuteron-induced reactions in combination with some theoretical models. In our previous works, the code system was successfully applied to systematic analyses of double-differential cross sections (DDXs) of (d,xp ) reactions for 12 C, 27 Al, and 58 Ni at incident energies up to 100 MeV. In the present work, we apply the code system to neutron emission from deuteron-induced reactions. Since there is few experimental data of DDXs of (d,xn ) reactions, double-differential thick target neutron yields (TTNYs) are calculated and compared with experimental data instead of DDXs. The calculation using the code system reproduces the measured TTNYs for carbon at incident energies up to 50 MeV.

Graphical imaging system for shifted spectrum in daughter nuclei of charged-current v-nucleus reactions

In this work, we improve the automated algorithm derived previously for drawing the quantum energy levels, in order to be efficient for plotting the theoretical spectra of the daughter nuclei in charged-current neutrino-nucleus reactions. Such calculated spectra need to be shifted so as the charged-current transition energies to be measured from the ground state of the daughter nucleus. The improvement involves the development of a computational code and graphical imaging system able to treat appropriately and as accurate as possible the energy-shifting which is different for the various multipole sets of nuclear states. The method is applied in the case of antineutrino-nucleus reaction which is of current experimental interest.

Systematic investigation of spectroscopic factors from (d,p) reactions for deuteron nuclear data evaluation

Spectroscopic factors for the (d,p) reactions on 12C, 27Al, 40Ca, and 58Ni for incident deuteron energies up to 100 MeV were extracted systematically by fitting theoretical distorted wave Born approximation calculations to the existing experimental data. The extracted spectroscopic factors show similar incident energy dependence for all the target nuclei. An empirical expression describing the energy dependence has been deduced and implemented in a computational code system developed specially for deuteron-induced reactions. The calculation with the code system reproduces the measured 27Al(d,p)28Al cross sections in the incident energy range from the threshold to 20 MeV and the measured double-differential cross sections for (d,xp) reactions on 12C and 58Ni at 56 and 100 MeV fairly well.

Development of a Calculation Code System for Evaluation of Deuteron Nuclear Data

Abstract A calculation code system for evaluation of deuteron nuclear data is extended so that the stripping reaction to bound states in the residual nucleus can be taken into account properly using a conventional zero-range DWBA approach. The code system is applied to deuteron induced-reactions on 27 Al for incident energies up to 100 MeV. It is found that the spectroscopic factors derived from the present DWBA analysis have incident energy dependence. The calculation using the extended code system reproduces experimental double-differential cross sections for the 27 Al(d,xp) reaction at 25.5, 56, and 100 MeV, and production cross sections of 28 Al in the incident energy range from the threshold to 20 MeV.

Computational analysis of neutronic parameters for TRIGA Mark-II research reactor using evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3

Important kinetic parameters such as effective multiplication factor, keff, excess reactivity, neutron flux and power distribution, and power peaking factors of TRIGA Mark II research reactor in Bangladesh have been calculated using the comprehensive neutronics calculation code system SRAC 2006 with the evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3. In the code system, PIJ code was employed to obtain cross section of the core cells, followed by the integral calculation of neutronic parameters of the reactor conducted by CITATION code. All the analyses were performed using the 7-group macroscopic cross section library. Results were compared to the experimental data, the safety analysis report (SAR) of the reactor provided by General Atomic as well as to the simulated values by numerically benchmarked MCNP4C, WIMS-CITATION and SRAC-CITATION codes. The maximum power densities at the hot spot were found to be 169.7W/cc and 170.1W/cc for data libraries ENDF/B-VII.0 and JENDL-3.3, respectively. Similarly, the total peaking factors based on ENDF/B-VII.0 and JENDL-3.3 were calculated as 5.68 and 5.70, respectively, which were compared to the original SAR value of 5.63, as well as to MCNP4C, WIMS-CITATION and SRAC-CITATION results. It was found in most cases that the calculated results demonstrate a good agreement with our experiments and published works. Therefore, this analysis benchmarks the code system and will be helpful to enhance further neutronics and thermal hydraulics study of the reactor.

Bounding Parameters of a Theoretical Deep-Earth Uranium Fission Reactor

More than 50 years ago it was postulated that a naturally occurring nuclear reactor was possible if it started 2 billion years ago. The subsequent discovery of the natural reactor at Oklo confirmed that it is possible for a nuclear fission reactor to naturally form and cycle on and off over extended periods of time. The hypothesis of a naturally occurring reactor was extended to include the possibility of significant quantities of uranium aggregating inside the molten core of Earth due to gravity during its formation. When sufficient quantities of uranium accumulate, a self-sustaining fission reactor could form, which would fluctuate in power as uranium fissions and fission products are produced. Lighter elements would migrate out of the reactor region, and heavier elements would coalesce due to gravity. In this technical note, SCALE, a state-of-the-art nuclear engineering computer code system developed at Oak Ridge National Laboratory, was used to investigate this hypothesis.The analysis indicates that the overall operational parameters of a postulated nuclear fission reactor located in the inner core of Earth must fall within a relatively narrow band in order to still be operating today. If the overall power level were too low, the reactor would not breed sufficient fissile material, and the average enrichment would drop below the level required to form a self-sustained fast reactor. If the power level were too high, the reactor would have burned itself out well before the present day. The objective of this technical note is to provide calculations that support an existing geo-reactor and the operating parameters that would govern a deep-Earth reactor and allow it to still be operating today, 4.5 billion years after Earth was formed. To help bound the possible power range, a simplified, one-dimensional, homogeneous, deep-Earth reactor having a steady-state fission power is simulated over geologic time. Power levels and start times are varied. The simulations show that if the reactor were still operating today, it would have an overall lifetime average operating fission power of <3 TW. Analyses show that both instantaneous and cumulative 3He/4He ratios are a function of fission power, 235U/238U ratio, total uranium mass, and geo-reactor starting time. The calculated 3He/4He ratios are consistent with those observed in nature.

Numerical modelling of radionuclide inventory for RBMK irradiated nuclear fuel

After the final shutdown of Ignalina NPP Unit 1 in 2004 and Unit 2 in 2009, total amount of spent nuclear fuel (SNF) is approximately 22,000 of fuel assemblies. Radionuclide content in SNF and its characteristics are initial data for analysis of various subjects such as radiation shielding, thermal analysis and other processes that can affect the performance of SNF storage or disposal facilities’ components. For instance, ageing of the components is influenced by decay heat, neutron and gamma radiation, external environment conditions, mechanical stresses, etc. The first two factors can be defined only after SNF characteristics are investigated and irradiation levels are determined. Experimental investigation of radionuclide content and characteristics in SNF is expensive and complicated, so numerical methods are widely used. In this paper, results of numerical modelling of spent RBMK-1500 nuclear fuel characteristics and its variation with time are presented. Modelling was performed using SCALE computer code system TRITON module. Some calculation results are compared with available experimental data for RBMK-1000 spent nuclear fuel and a rather good agreement between numerical and experimental results is identified. Comparison of calculated gamma and neutron sources, activities and decay heats shows that results for RBMK-1500 fuel with various enrichments differ by 10–20%.

Automated drawing system of quantum energy levels

The purpose of this work is to derive an automated system that provides advantageous drawings of energy spectra for quantum systems (nuclei, atoms, molecules, etc.) required in various physical sciences. The automation involves the development of appropriate computational code and graphical imaging system based on raw data insertion, theoretical calculations and experimental or bibliographic data insertion. The system determines the appropriate scale to depict graphically with the best possible way in the available space. The presently developed code operates locally and the results are displayed on the screen and can be exported to a PostScript file. We note its main features to arrange and visualize in the available space the energy levels with their identity, taking care the existence in the final diagram the least auxiliary deviations. Future improvements can be the use of Java and the availability on the Internet. The work involves the automated plotting of energy levels in molecules, atoms, nuclei and other types of quantized energy spectra. The automation involves the development of an appropriate computational code and graphical imaging system.

Reconstruction of cell homogenized macroscopic cross sections for analyzing fast and thermal coupled cores using the SRAC system

A fast and thermal neutron coupled core adopts blanket fuel assemblies with zirconium hydrides in the core for negative coolant void reactivity. Conventional neutronics calculation methods have been developed for analysis of a fast core or thermal core, in which the coarse-group macroscopic cross sections of fuel assemblies are prepared without including the effect of the surrounding fuel assemblies. However, such methods are not adequate for analyzing fast and thermal neutron coupled cores where the intra-assembly and inter-assembly heterogeneity effects must be precisely taken into account. Recently, a concept of reconstruction of cell homogenized macroscopic cross sections has been proposed to take into account effects of inter-assembly heterogeneities on macroscopic cross sections used in the reactor core analysis and successfully applied based on a Monte Carlo method. In the present study, a reconstruction method of cell homogenized coarse-group macroscopic cross section for analyzing fast and thermal coupled cores is developed based on a deterministic neutronics calculation code system, SRAC. Three types of fixed source calculations for unit assembly cell geometry are performed independently of the specific core layouts and their results are combined with the results of core analysis to produce cell homogenized coarse-group macroscopic cross sections. Numerical results show that the heterogeneity effects can be adequately reflected in the reconstructed macroscopic cross sections with the proposed method. When the number of energy groups is small, the proposed method gives poor results in the transitional energy groups from resonance to thermal energy. Therefore, it is necessary to increase the number of energy groups in this energy range.

Depletion Reactivity Benchmarks—I: Experimental Benchmarks for Quantifying PWR Fuel Reactivity Depletion Uncertainty

Pressurized water reactor (PWR) assembly reactivity distributions are inferred from ~600 in-core flux maps taken during 44 cycles of operation of the Catawba and McGuire nuclear power plants. The reactivity distribution for each flux map is determined by systematically searching for fuel subbatch reactivities that minimize differences between measured and computed 235U fission rates. More than eight million core calculations are used to reduce one million measured signals to a set of ~2500 experimental fuel reactivities for fuel with up to 55 GWd/T burnup. These measured reactivity changes with depletion can be used to validate computer code systems used for burnup credit.To reduce the effort required to quantify computer code system biases and uncertainties, the measured changes in fuel depletion reactivity have been reduced to a set of experimental PWR lattice benchmarks for the change in reactivity as a function of fuel burnup. Results demonstrate that the uncertainty of hot-full-power (HFP) depletion reactivity of the benchmarks is < 250 pcm up to 55 GWd/T burnup. Oak Ridge National Laboratory’s TSUNAMI tools are used to extend HFP results to cold conditions, and reactivity decrement uncertainties increase to ~600 pcm. These experimental benchmarks provide a basis for quantification of combined nuclide inventory and cross-section uncertainties in computed reactivity decrements.It is demonstrated that flux map data reduction is not sensitive to the analytical tools (CASMO/SIMULATE) employed here, and experimental fuel depletion reactivity decrements and uncertainties are anticipated to be independent of fuel management code system use for the data reduction. For CASMO-based analysis, the HFP reactivity burnup decrement biases are shown to be <250 pcm up to 55 GWd/T burnup, and results show that the historical “Kopp memo” 5% reactivity decrement uncertainty assumption is conservative.

Photon transport effect on intra-subassembly thermal power distribution in fast reactor

In order to accurately predict intra-subassembly thermal power distribution in a fast reactor, neutron and photon transport calculations are carried out with a multi-purpose reactor physics calculation code system CBZ. All the fission fragment nuclide are treated explicitly during fuel depletion, and irradiation time-dependent energy spectra of delayed fission γ-rays emitted from all the fission fragment nuclides are precisely simulated. Time-dependent delayed β-ray emission and transmutations of fission fragment nuclide by neutron-nuclide reactions are also taken into account. A fuel subassembly model of Japanese prototype fast reactor Monju is used for numerical calculations, and their two-dimensional geometric feature is precisely modeled by a ray-tracing-based collision probability method implemented in CBZ. When the photon transport is considered, total thermal powers in fissile material regions are reduced by about 1.5% except at the beginning of fuel depletion.

Review of studies on criticality safety evaluation and criticality experiment methods

Since the early 1960s, many studies on criticality safety evaluation have been conducted in Japan. Computer code systems were developed initially by employing finite difference methods, and more recently by using Monte Carlo methods. Criticality experiments have also been carried out in many laboratories in Japan as well as overseas. By effectively using these study results, the Japanese Criticality Safety Handbook was published in 1988, almost the intermediate point of the last 50 years. An increased interest has been shown in criticality safety studies, and a Working Party on Nuclear Criticality Safety (WPNCS) was set up by the Nuclear Science Committee of Organisation Economic Co-operation and Development in 1997. WPNCS has several task forces in charge of each of the International Criticality Safety Benchmark Evaluation Program (ICSBEP), Subcritical Measurement, Experimental Needs, Burn-up Credit Studies and Minimum Critical Values. Criticality safety studies in Japan have been carried out in cooperation with WPNCS. This paper describes criticality safety study activities in Japan along with the contents of the Japanese Criticality Safety Handbook and the tasks of WPNCS.

Development of a MCNP–ORIGEN burn-up calculation code system and its accuracy assessment

An MCNP–ORIGEN burn-up calculation code system, named MCORE (MCNP and ORIGEN burn-up Evaluation code), is developed in this work. MCORE makes use of the Monte Carlo neutron and photon transport code MCNP4C and nuclides depletion and decay calculation code ORIGEN2.1. MCNP and ORIGEN are coupled by data processing and linking subroutines. In MCORE, a so called “modified predictor corrector” approach is used. MCORE provides the capability of using different depletion calculation schemes and simulating fuel shuffling. Total nuclide density changes in active cells are considered in MCORE. The validity and applicability of the developed code are tested by investigating and predicting the neutronic and isotopic behavior of a “VVER-1000 LEU Assembly Computational Benchmark” at lattice level and a “Physics of Plutonium Recycling” fast reactor at core level (OECD-NEA).The comparison results show that the MCORE code predicts the nuclide composition within 5% accuracy and k∞ within 800pcm at the end of the burn-up for LEU assembly (40MWD/kg HM). For a fast reactor, the results obtained by MCORE are in the range of reported results except for 243Am. In general, MCORE results show a good agreement with the benchmark values.

Analysis of kinetic parameters of 3 MW TRIGA Mark-II research reactor using the SRAC2006 code system

The aim of this study is to evaluate the kinetic parameters of 3MW TRIGA Mark-II research reactor at AERE, Savar, Dhaka, Bangladesh from the viewpoint of reactor safety. The most important kinetic parameters of nuclear reactors are the effective delayed neutron fraction (βeff), the effective decay constant for ith family of delayed neutron precursor (λeff,i), the prompt neutron lifetime (lp) and the mean neutron generation time (Λ). These parameters are calculated using the 3-D diffusion code SRAC-CITATION of the comprehensive neutronics calculation code system SRAC2006 based on the evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 in both cases. The calculated results of reactor kinetic parameters are compared to the available safety analysis report (SAR) values of 3MW TRIGA Mark-II reactor by General Atomic as well as the MCNP5 values (numerically benchmark) based on the evaluated nuclear data library ENDL/B-VII.0. It was found that in most cases, the calculated results of kinetic parameters demonstrate a good agreement between the JENDL-3.3 and the ENDF/B-VII libraries as well as the SAR and the MCNP5 values respectively. Therefore, this study will be essential to improve the basic nuclear data of reactor kinetic parameters for safe operation of 3MW TRIGA Mark-II research reactor.

Neutronic characterization and decay heat calculations in the in-vessel fuel storage facilities for MYRRHA/FASTEF

The main objective of the Central Design Team (CDT) project is to establish an engineering design of a Fast Spectrum Transmutation Experimental Facility (FASTEF) that is the pilot plant of an experimental-scale of both an Accelerator Driven System (ADS) and a Lead Fast Reactor (LFR), based on the MYRRHA reactor concept, planned to be built during the next decade. The MYRRHA reactor concept is devoted to be a multi-purpose irradiation facility aimed at demonstrating the efficient transmutation of long-lived and high radiotoxicity minor actinides, fission products and the associated technology. An important issue regarding the reactor design of the MYRRHA/FASTEF experiment is the In-Vessel Fuel Storage Facilities (IVFSFs), both for fresh and spent fuel, as it might have an impact on the criticality of the overall system that must be quantified. In this work, the neutronic analysis of the in-vessel fuel storage facility and its coupling with the critical core was performed, using the state of the art Monte Carlo program MCNPX 2.6.0 and ORIGEN 2.2 computer code system for calculating the buildup and decay heat of spent fuel. Several parameters were analyzed, like the criticality behavior (namely the Keff), the neutron fluxes and their variations, the fission power production and the radiation damage (the displacements per atom). Finally, also the heat power generated by the fission products decay in the spent fuel was assessed.

Shutdown dose rate calculation code system and its application to EAST tokamak

For safe operation and maintenance of nuclear devices, it is very important to predict the dose rate distribution after shutdown. Based on the rigorous two-step (R2S) method, a new shutdown dose rate calculation code system was developed for nuclear devices with large dimension and complex geometries. By coupling the Monte Carlo particle transport calculation code MCNP with the activation simulation code FISPACT, the dose rate calculation codes integrate the functions of neutron transport calculation, activation calculation and decay gamma transport calculation. This code system was applied to the shutdown dose analysis for experimental advanced superconducting tokamak (EAST). Three-dimensional dose rate distributions of the EAST for different cooling times and spatial locations were obtained. The results can be used to assist radiation protection in EAST.