Nuclear Libraries Research Articles

Development and verification of neutron and photon multi-group cross sections processing code TXMAT2.0

TXMAT2.0 is an in-house code for processing neutron and photon cross sections from the MATXS format nuclear library, which provides multi-group macroscopic cross sections for transport and diffusion calculations. In developing the program, three improvements have been made in the code. First, TXMAT2.0 adopts Fortran2003 based on object-oriented programming techniques and uses allocatable memory management techniques. Second, TXMAT2.0 is capable of processing the ultrafine group (2082 groups) library using these memory management techniques; in comparison, the officially released TRANSX2.15 has difficulty with this task. Finally, TXMAT2.0 handles correctly self-shielding corrections for light nuclides such as 12C, 16O, whereas TRANSX 2.15 experiences problems. In performing the verification and validation of TXMAT2.0, ICSBEP benchmarks for criticality verification, and OKTAVIAN and IPPE benchmarks for shielding verification were selected. Several results were compared with results obtained from TRANSX2.15 as well as with experimental data. The results show that the accuracy of TXMAT2.0 is as good as TRANSX2.15. Moreover, TXMAT2.0 provides accurate neutron and photon cross sections for multi-group transport calculations. In addition, through the verification of a self-designed benchmark, TXMAT2.0 processes effectively and correctly self-shielding problems of light nuclides.

Nuclear data processing code FRENDY: A verification with HTTR criticality benchmark experiments

Japan Atomic Energy Agency has developed a new nuclear data processing code, named FRENDY, to generate ACE-formatted files from various evaluated nuclear libraries. A code-to-experiment verification of FRENDY processing was carried out in this study with criticality benchmark assessments of the high temperature engineering test reactor. The ACE files of the JENDL-4.0 and ENDF/B-VII.1 was generated successfully by FRENDY. These ACE files have been used in MCNP6 transport for various benchmark problems of the high temperature engineering test reactor. As a result, the keff and reaction rate obtained by MCNP6 calculation presented a good agreement compared to the experimental data. The proper ACE files generation by FRENDY was confirmed for the HTTR criticality calculations.

Nuclear performances of the water-cooled lithium lead DEMO reactor: Neutronic analysis on a fully heterogeneous model

The development of a conceptual design for the Demonstration Fusion Power Reactor (DEMO) is a key issue within the EUROfusion roadmap. The DEMO reactor is designed to produce a fusion power of about 2 GW and generate a substantial amount of electricity, relying on a closed tritium fuel cycle: it implies that the breeding blanket (BB) shall guarantee a suitable tritium production to enable a continuous operation without any external supply.The Water-Cooled Lithium Lead (WCLL) concept is a candidate for the DEMO BB: it uses liquid Lithium Lead as breeder and neutron multiplier and water in PWR condition as coolant. The neutronics analyses carried out in the past have been performed using a semi-heterogeneous representation of the BB, since the complexity of its structure makes the generation of a detailed MCNP model a very demanding and challenging task. Results highlighted good performances for the WCLL BB, both in terms of shielding effectiveness and tritium self-sufficiency.A recently updated assessment of the tritium breeding ratio (TBR) requirement for DEMO, considering margins for calculation uncertainties and incomplete models of the whole machine, led to the definition of a tentative 1.15 value for the TBR. Moreover, the implementation of an accurate BB neutronics model, consistent with the engineering design, is recommended for the evaluation of the tritium self-sufficiency. In order to tackle these issues, an MCNP model of the DEMO tokamak, integrating a fully heterogeneous WCLL BB has been developed for the first time, including an accurate description of the FW water channels, as well as a comprehensive definition of the breeding zone inner structure. A complete assessment of the WCLL BB nuclear performances, through 3D neutron and gamma transport simulations, has been carried out by means the MCNP Monte Carlo code and JEFF nuclear libraries.

Maxwellian-Averaged Neutron Capture Cross-Sections and Thermonuclear Reaction Rates for 56,57,58Fe, 59Co, And 60Ni Isotopes at Astrophysical Energies

Initially-produced isotopes during the course of the s-process have a significant impact on the continuity and branching network of this process. Such isotopes can undergo various neutron capture mechanisms, of which (n, γ) is important. In this research, the direct, thermal, and Maxwellian-averaged cross section (MACS), as well as the astrophysical reaction rates of the radiative neutron-capture reactions, were calculated for 56,57,58Fe, 59Co, and 60Ni isotopes. At kinetic energies between kT = 0.037 and 482.3 keV, corresponding to astrophysical temperature in the range of 0.01 to 1GK. E1 transitions of s-wave were considered with the approximations of direct non-resonant components. The obtained MACSs and hence reaction rates were based on the data from the major measured and evaluated nuclear libraries and tables and they showed very well agreements with the most recommended database and compilations. Moreover, the effect of the direct, thermal, and negative resonance cross sections on the MACS was found to be very small and exhibit a 1/v behavior overall temperatures, with a maximum at T = 0.01GK, while the main contribution attributed to the positive resonance term, that found to dominate overall temperature range.

CRITICAL EXPERIMENT OF THORIUM LOADED THERMAL CORES AT KUCA (1) A NEW CRITICAL EXPERIMENT OF THORIUM LOADED CORE WITH HARDER NEUTRON SPECTRUM IN KUCA

In order to perform integral evaluation of 232Th capture cross section, a series of critical experiments for thorium-loaded and solid-moderated cores in KUCA had been carried out. In these experimental cores, H/235U nuclide ratio ranged about from 150 to 315, and 232Th/235U nuclide ratio ranged about from 13 to 19. In this study, a new critical experiment with Th loaded core in KUCA, which had about 70 of the H/235U ratio and 12.7 of 232Th/235U ratio, was carried out. As results, the excess reactivity was 0.086 ± 0.003 (% dk/k) and the keff was 1.0009 ± 0.0003, where the effective delayed neutron fraction was 7.656E-3. The keff was also calculated by MVP3.0 with different nuclear libraries. The respective calculations with JENDL-4.0, JENDL-3.3 and ENDF/B-VII.0 lead to 1.0056 ± 0.0086 (%), 1.0048 ± 0.0085 (%) and 1.0056 ± 0.0086 (%).On the other hand, the further MVP3.0 calculations, where only the 232Th cross sections were taken from JENDL-4.0, JENDL-3.3 or ENDF/B-VII.0 but all other nuclides were done from JENDL-4.0, were carried out to examine an impact of the difference of 232Th cross section among these nuclear libraries to the keff. The keff calculated with respective 232Th cross sections from JENDL-3.3 and ENDF/B-VII.0 was 1.0038 ± 0.0086 (%) and 1.0040 ± 0.0086 (%).

FEATURES OF THE SOURCE TERM CALCULATION CAPABILITY OF THE NECP-X CODE

The source term calculation capability is developed for the high-fidelity neutronics code NECP-X. Generally, a full activation library is used, but the memory requirement is unacceptable for the high-fidelity calculation. In order to minimize the memory requirement during the calculation with very strict conditions, a new generalized activation chain compressed method is proposed based on the influence qualification. One basic compression element is a reaction channel or an isotope, and the influence of every compression element to the final results are qualified. To enlarge the range of application of the new compressed library, an effective method to determine representative problems, which utilizes the neutron spectra and neutron flux, is developed and analyzed. Based on the ENDF-VII.0, EAF-2010 evaluated nuclear library and the influence qualification-based activation library compression method, a new compressed activation library is generated. The VERA-3A problem and the KAIST problem are used to assess the accuracy and the efficiency of the new activation library. 85 measurements of decay heat from decay heat measurement facilities GE-Morris and CLAB are used to validate the decay heat calculation in NECP-X. The results show good accuracy of NECP-X in predicting radiation source term of the spent nuclear fuel and significant memory saving when using new compressed activation library.

JADE, a new software tool for nuclear fusion data libraries verification & validation

• Automatic verification and validation of nuclear data libraries. • Benchmarks for cross sections verification. • Fusion related nuclear data libraries verification and validation. • Spotting inconsistencies and errors in cross sections. Nuclear data improvement is a key objective in the nuclear field and, in particular, for fusion application. Due to their complex production process, new nuclear library releases must undergo systematic and extensive Verification and Validation (V&V) procedures. A novel software tool, JADE, is presented in this work. Its aim is to lead the path to a full standardization and automatization of V&V processes for neutron-gamma Monte Carlo transport for fusion applications, as well as saving substantial user time and avoid error-prone editing operations. JADE is a Python 3 based software that will be able to automatically run and post-process an exhaustive suite of computational and experimental benchmarks, together with the incorporation of pre-existing quality checks strategies. A first computational benchmark, the Leakage Sphere benchmark, has been implemented, a series of nuclear libraries were tested with it and the results have been discussed. The capability of JADE for spotting anomalies in the cross sections has been conclusively proven by being able to highlight inconsistencies already documented in the related literature. Moreover, some previously unknown anomalies have been identified and discussed, giving evidence of JADE’s applicability for the validation of future nuclear libraries releases. Finally, future developments and objectives are discussed.

Design Study of Gas Cooled Fast Reactor (GFR) with Uranium Plutonium Carbide (UC-PuC) as Fuel with Addition Protactinium (Pa-231)

Analysis performance of uranium plutonium carbide (UC-PuC) as fuel in gas cooled fast reactor (GFR) with addition of protactinium as a burnable poisons has been done. Neutronic analysis in this research was carried out using the SRAC code from JAERI with a nuclear library based on JENDL 4.0. The calculation is carried out by two steps, the first step is the PIJ calculation which calculates the fuel cell and the second step is the CITATION calculation which calculates the various configurations of the reactor core. The first calculation determines the k-eff value in a homogeneous core configuration. The results obtained show that the percentage of 10% is the sloping result with a k-eff value of 1%. The second calculation determines the k-eff value in the heterogeneous core configuration. The results obtained indicate that the fuel variation 8% -10% -12% is the most critical percentage with a peak power density value of less than 100 Watt/cc. Furthermore, the addition of protactinium with a variation of 0% to 5%. At a protactinium 4% percentage and 63% fuel fraction, the excess reactivity value is 1.02% or close to 1% which indicates that the reactor is in a critical condition.

A new generalized activation chain compression method in NECP-X

The full-fidelity activation library is generally used for the activation photon source term calculation, but the memory requirement is unacceptable for the high-fidelity calculation. In order to minimize the memory requirement, a new generalized activation chain compression method is proposed. One basic compression element is a reaction channel or a decay path of an isotope, and the influence of every compression element to the final results are qualified. To enlarge the range of application of the new compressed library, an effective method to determine representative problems is developed and analyzed. Based on the ENDF-VII.0, EAF-2010 evaluated nuclear library and influence qualification-based activation library compression method, a new compressed activation library is generated. A set of problems are tested, including 2D VERA assembly problems, the VERA-3A problem and the KAIST problem, to assess the accuracy and the efficiency of the new activation library, showing its good accuracy and significant memory saving.

Design Conceptual of 800MWt Long Life Pressurized Water Reactor Using (Th-U)O2 Fuels with Gd2O3 and Pa-231 as Burnable Poisons

A long-life pressurized water reactor (PWR) has been reviewed as an innovative reactor design that can fulfill electricity demand. This study aimed to find out the optimum design of 800MWt long life PWR using Thorium-Uranium dioxide (Th-U)O2 fuels with Gadolinium (Gd2O3) and Protactinium-231 (Pa-231) as the burnable poisons. An established computer code of SRAC 2006 with JENDL 4.0 as data nuclear library had been used for the analysis. A two-dimensional R-Z geometry and fuel volume fraction of 40% were used for core geometry analysis. The different fraction of Uranium dioxide, Uranium-235, Gadolinium, and Protactinium-231 had been carried out. The result of this study was a design of PWR 800MWt using Uranium dioxide fuel of 60% with enrichment 11%-12%-13% Uranium-235 and the addition of 0,025% Gd2O3 and 1,0% Pa-231 that could operate for ten years without refueling. The reactor could produce a power density of 45,4 watts/cc with excess reactivity about 3,6% dk/k. This study is expected to be a reference for a long-life pressurized water reactor using the Thorium-Uranium fuel cycles.

Nuclear Data Covariance Analysis in Radiation-Transport Simulations Utilizing SCALE Sampler and the IRDFF Nuclear Data Library

This article describes the nuclear data covariance analysis of an experimental design for a neutron energy-tuning assembly (ETA) created to shape a 14-MeV neutron point source to an objective spectrum. Underlying nuclear data uncertainties play a large role in the radiation transport and reaction rates for the range of responses to be expected from an experiment. The methodology leveraged the Standardized Computer Analysis for Licensing Evaluation (SCALE) Sampler module to determine the uncertainty in the neutron transport. The reaction uncertainty was perturbed with the International Reactor Dosimetry and Fusion File v.1.05 uncertainty, correlation matrix, and reaction cross section through multivariate normal distribution sampling to provide a final response metric. The resultant neutron fluence uncertainty for the ETA ranged from 2.7% to 6.2% in the energy range from 1.28 keV to 14.1 MeV, which contains 99.99% of the neutron fluence. The integrated uncertainties, including statistical and systematic nuclear data uncertainties, for the reaction products analyzed were 2.33% to 4.84% for most reactions, but 55Mn(n, $\gamma $ ), a less well-characterized reaction occurring in an energy domain with high flux uncertainty, was 19.7%. The mean of the reaction distributions was within 1.1% of the unperturbed nuclear data simulation. The experiment is planned for late 2019, where the predicted results will be compared against the experimental outcomes. The methodology presented can be utilized with alternate nuclear libraries in SCALE to develop uncertainty bounds and neutron flux spectra for many radiation-transport problems.

Benchmarking the new ENDF/B-VIII.0 nuclear data library for OECD/NEA medium 1000 MWth sodium-cooled fast reactor

This paper presents the benchmark evaluation of the new ENDF/B-VIII.0 nuclear library for the OECD/NEA Medium 1000 MWth Sodium-cooled Fast Reactor (SFR). There are 2 SFR cores: metallic fueled (MET-1000) and oxide fueled (MOX-1000). The continuous-energy Monte Carlo Serpent2 code was used as the calculation tool. Various nuclear libraries such as ENDF/B-VII.1 and JENDL-4.0 were included to be compared with the newest ENDF/B-VIII.0. The evaluated parameters are k,βeff, sodium void reactivity (∆ρNa), Doppler constant (∆ρDoppler), and control rod worth (∆ρCR).

Interpretation of two SINBAD photon-leakage benchmarks with nuclear library ENDF/B-VIII.0 and Monte Carlo code MCS

A review of the documentation and an interpretation of the NEA-1517/74 and NEA-1517/80 shielding benchmarks (measurements of photon leakage flux from a hollow sphere with a central 14 MeV neutron source) from the SINBAD database with the Monte Carlo code MCS and the most up-to-date ENDF/B-VIII.0 neutron data library are conducted. The two analyzed benchmarks describe satisfactorily the energy resolution of the photon detector and the geometry of the spherical samples with inner beam tube, tritium target and cooling water circuit, but lack information regarding the detector geometry and the distances of shields and collimators relatively to the neutron source and the detector. Calculations are therefore conducted for a sphere model only. A preliminary verification of MCS neutron-photon calculations against MCNP6.2 is first conducted, then the impact of modelling the inner beam tube, tritium target and cooling water circuit is assessed. Finally, a comparison of calculated results with the libraries ENDF/B-VII.1 and ENDF/B-VIII.0 against the measurements is conducted and shows reasonable agreement. The MCS and MCNP inputs used for the interpretation are available as supplementary material of this article.

Reactivity feedback evaluation during the start-up of the heat pipe cooled nuclear reactors

Alkali metal heat pipe cooled reactors (HPRs) feature a compact structure, high power capacity, and longer lifetime than conventional chemical or solar power sources. Thus, HPRs could be used as energy source for long-term space missions, such as deep space or planetary surface exploration. The reactivity temperature coefficient (RTC) is a key parameter for reactor safety, and RTC for HPRs is influenced by multiple factors. The impact mechanism of the heat pipes working fluid on the reactivity temperature feedback for HPRs consists of two aspects: 1) Metal vapor status in the internal space; 2) Height variation of the liquid pool at heat pipe bottom. In this study, a HPR model is established consisting of 90 uranium nitride (UN) fuel pins and 37 heat pipes, a BeO reflector and 6 control drums with B4C absorbers. Different types of heat pipe working fluid are simulated, including sodium and potassium. The MCNP code and the ENDF-B-VII.0 nuclear library are used for the calculation. Numerical results show that the metal vapor in the internal space of the heat pipes provides a positive feedback, while RTC provided by the liquid pool at heat pipe bottom is not monotonous. At the heat pipe operation temperature, RTC provided by the heat pipes are 0.08 pcm/K for Na HPR and 0.06 pcm/K for K HPR. Considering all the influential factors, the RTC for HPRs is negative (-0.83 pcm/K for Na and -0.91 pcm/K for K) and the HPR features strong self-stabilizing capability. This study could provide a reference for the design of HPRs.

Evaluation of displacement cross-section for neutron-irradiated 15-15Ti steel and its swelling behavior in CiADS radiation environment

We report on the proposed method of evaluating the displacement cross-sections in NRT and BCA-arc-dpa models. The method is based on the NJOY code in conjunction with the IOTA code and is applicable to neutron energy in the range of 0.1 neV–20 MeV, defined by using NJOY and nuclear library files with data up to 20 MeV. The displacement cross-section has been calculated for natural iron, 15-15Ti, 316L, Eurofer, ODS, and T91 steels. The displacement cross-sections for 15-15Ti steel were used to estimate radiation damage and swelling level under irradiation conditions of the China Initiative Accelerator Driven System (CiADS) subcritical core. The swelling analysis has shown, that in CiADS radiation environment, the 15-15Ti steel has a good margin to cladding failure due to swelling and, therefore, might be a cladding material for CiADS.

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.

Comparison of first criticality prediction and experiment of the Jordan research and training reactor (JRTR)

Abstract Korea Atomic Energy Research Institute (KAERI) has carried out various neutronics experiments in the commissioning stage of the Jordan Research and Training Reactor (JRTR), and this paper introduces the results of first criticality prediction and experiment for the JRTR. The Monte Carlo Code for Advanced Reactor Design and analysis (McCARD) with the ENDF/B-VII.0 nuclear library was used for prediction calculations in the process of the first criticality approach, which was performed to provide reference for the first criticality experiment. In the experiment, fuel loading was carried out by measuring the inverse multiplication factor (1/M) to predict the number of fuel assemblies at the first criticality, and the first critical was reached on April 25, 2016. Comparing the first criticality prediction and experiment, the calculated and measured CAR (Control Absorber Rod) heights for the first criticality were 575 mm and 570.5 mm, respectively, that is, the difference between the two results was approximately 5 mm. From this result, it was confirmed that JRTR manufacturing and various experiments had successfully progressed as designed.

Analysis of VIP-BWR reactor core physics experiments on UO2 and MOX mockup fuel assemblies with CASMO5

ABSTRACT The VIP-BWR program comprises reactor core physics experiments on three core configurations each consisting of BWR mockup, reference, and driver fuel assemblies. The mockup fuel assemblies are an all-UO2, an island-type MOX and a full-MOX-type 8 × 8 fuel assembly each including Gd2O3-UO2 fuel rods. The measurement results include the critical water levels and fission rates for large numbers of fuel rods. Analysis of the experiments was performed in a two dimensional geometrical model with the lattice physics code CASMO5 coupled with the JENDL-4.0-based nuclear library. The axial leakage of neutrons from the core was taken into account adopting the measured axial buckling. The effective multiplication factors of the critical cores ranged from 0.9940 to 0.9976, with the measurement uncertainties of 0.0003 to 0.0004 (1 σ). The calculated results of core radial fission rate distributions were compared with the measurements. For the cores loaded with the MOX fuel assemblies, the calculations overestimated the fission rates of MOX fuel rods by 4 to 5% relatively to those of UO2 and Gd2O3-UO2 fuel rods. It agreed with the trends in the other preceding analysis studies, and biases were confirmed in the calculation results of power distributions in MOX-fueled light water reactor cores.

Benchmarking of evaluated nuclear data for iron by a TOF experiment with slab samples

In order to validate the evaluated nuclear data of iron, neutron leakage spectra in the range from 0.8 to 15 MeV from natural iron slab samples were measured by time-of-flight technique at 60° and 120°, the measurements were performed at China Institute of Atomic Energy (CIAE) using a D-T neutron source, the slab thicknesses were chosen to be 5 cm, 10 cm and 15 cm, corresponding to 1.08, 2.16 and 3.24 mean free paths for 14.5 MeV neutrons, respectively. Theoretical calculations were carried out by Monte Carlo neutron transport code MCNP-4C, using the evaluated nuclear libraries of iron from the CENDL-3.1, ENDF/B-VIII.0, JENDL-4.0, JEFF-3.3, and TENDL-2017. The experimental results were compared with the corresponding calculated ones. From the comparison, it was found that: (1) In the 12–15 MeV range, the results with the ENDF/B-VIII.0 library well reproduce the experimental ones at both 60°and 120°, but those with the JENDL-4.0, JEFF-3.3 and TENDL-2017 libraries are underestimated, especially at 120°; (2) In the 8.5-12 MeV energy range, the calculated neutron spectra with the TENDL-2017 library are underestimated by more than 35%, but those with the JENDL-4.0 library are overestimated around 20%.

Measurement of kinetic parameter of the Jordan research and training reactor

The kinetic parameter of the Jordan Research and Training Reactor (JRTR) was measured during commissioning. Neutron noise analysis techniques of correlation, power spectral density (PSD), and variance to mean ratio (VTMR) were applied to determine the prompt neutron decay constant (α) at several zero-power subcritical states. The ratio of the effective delayed neutron fraction to prompt neutron reproduction time (βeff/Λ) corresponding to α at a critical state was estimated by a linear extrapolation of α values as a function of inverse count rate. The βeff/Λ value determined by using the time series neutron data from the BF3 detectors was 54.76–60.32 s-1 and was compared to the calculated one of 61.70 s-1 by the McCARD code with the ENDF/B-VII.0 nuclear library. The relative difference between the measurement and the calculation was within 13%, regardless of the noise analysis techniques.