Accuracy of criticality calculations made using current nuclear data libraries for MOX fuelled fast critical assemblies

Criticality calculations have been made for a set of ten mixed plutonium–uranium oxide (MOX) fuelled fast critical assemblies using the current nuclear data libraries, JEFF-3.1, JEFF-3.1.1, JENDL-3.3 and ENDF/B-VII.0. The results obtained using the different libraries are compared and conclusions drawn concerning the accuracy of criticality calculations made for MOX fuelled fast reactors.     

Detail analysis of fusion neutronics benchmark experiment for iron

In order to specify the best nuclear data on iron, the fusion neutronics benchmark experiment on iron at Japan Atomic Energy Agency (JAEA)/Fusion Neutronics Source (FNS) was analyzed in detail with MCNP-4C and the latest nuclear data libraries, JENDL-3.3, FENDL-2.1, JEFF-3.1 and ENDF/B-VII.0. As a result, totally the calculation result with ENDF/B-VII.0 agreed with the measurement best, except that it underestimated the measured neutron flux above 10 MeV with the depth. It was noted that the calculation result with JENDL-3.3 overestimated the measured neutrons below a few keV. Through the DORT calculations based on the iron data in ENDF/B-VII.0, it was found out that the first inelastic scattering cross-section data of ⁵⁷Fe in JENDL-3.3 caused the overestimation.     

Validation of the sodium void reactivity effect prediction using JEFF-3.1 nuclear data

Near 60 Na void experiments performed in the zero power reactors MASURCA (CEA-Cadarache) and ZPPR (Argonne West – Idaho) have been analyzed using JEFF-3.1 nuclear data and the ERANOS-2.1 (deterministic) and TRIPOLI-4 (Monte-Carlo) codes. Some comparative calculations have been performed also using either JEFF-3.1, ENDF/B-VII.0 or JENDL-3.3 nuclear data for ²³Na, as these three ²³Na evaluations show marked differences. The Na void experiments have been selected to cover spectral conditions ranging from the relatively hard flux in the outer zone of a small fast reactor to the relatively soft flux in the inner zone of a large fast reactor. For in-fuel Na void patterns, there is a good agreement between ERANOS and TRIPOLI computations, while the deterministic calculations significantly underestimate the leakage component for Na void patterns in fertile regions. The agreement between ERANOS-2.1 + JEFF-3.1 predictions and experimental values is excellent for in-fuel Na void patterns in MASURCA experiments, but a significant underestimation of the leakage component occurs for in-fuel Na void patterns in ZPPR. For fertile Na void patterns, there is a clear underestimation of the leakage component, quantitatively different for MASURCA and ZPPR experiments. Variations in ²³Na cross-section data also result in significant differences: ENDF/B-VII.0 and JENDL-3.3 nuclear data for ²³Na increase noticeably the predicted Na void worth values with respect to JEFF-3.1 data. The three ²³Na evaluations differ at high energy (>500 keV, and especially >2 MeV), and this stresses the need for accurate additional measurements in this energy range.     

Exact nuclear data uncertainty propagation for fusion neutronics calculations

Recently, we have presented an exact method (now called “Total Monte Carlo”) to propagate uncertainties of fundamental nuclear physics experiments, models and parameters to different types of criticality-safety benchmarks. We now show that such exact uncertainty calculations are directly relevant to the optimal and safe design of fusion reactors by applying this methodology to a series of fusion shielding benchmarks, namely those connected to the Oktavian, Fusion Neutronics Source and LLNL Pulsed Sphere experiments. Uncertainties on neutron and gamma leakage fluxes for 13 shielding benchmarks are obtained, in the mass range from ^natMg to ^natW. Uncertainties for cross-sections, angular distributions, single- and double-differential emission spectra, and gamma-ray production cross-sections are considered in this uncertainty propagation scheme.     

Analyses with latest major nuclear data libraries of the fission rate ratios for several TRU nuclides in the FCA-IX experiments

In the 1980s, a series of integral experiments was conducted in FCA-IX assemblies with systematically changed neutron spectra covering from the intermediate to fast ones. The experiments provide systematic data of central fission rates for TRU nuclides containing minor actinides, ²³⁷Np, ²³⁸Pu, ²³⁹Pu, ²⁴²Pu, ²⁴¹Am, ²⁴³Am, and ²⁴⁴Cm. Regarding the fission rate ratios relative to ²³⁹Pu, benchmark models had been recently developed for validation of nuclear data for the TRU's fission cross sections. In this paper, the latest major nuclear data libraries, JENDL-4.0, ENDF/B-VII.1, and JEFF-3.2, are compared on the benchmark models. For the libraries, the analyses by a Monte Carlo calculation code show obvious overestimations particularly for the fission rate ratios of ²⁴⁴Cm to ²³⁹Pu. Additionally, a large discrepancy about by 20% between the libraries is revealed for the fission rate ratio of ²³⁸Pu to ²³⁹Pu measured in the intermediate neutron spectrum. The causes of discrepancies are furthermore clarified by sensitivity analyses.     

Validation of Pb nuclear data by Monte Carlo analyses of sample reactivity experiments at Kyoto University Critical Assembly

Sample reactivity experiments on the uncertainty analyses of Pb nuclear data are carried out by substituting Al plates for Pb ones at the Kyoto University Critical Assembly, as part of basic research on Pb–Bi for the coolant. Numerical simulations of sample reactivity experiments are performed with the Monte Carlo calculation code MCNP6.1 together with four nuclear data libraries JENDL-3.3, JENDL-4.0, ENDF/B-VII.0 and JEFF-3.1, to examine the accuracy of cross-section uncertainties of Pb isotopes by comparing measured and calculated sample reactivities. A library update from JENDL-3.3 to JENDL-4.0 is demonstrated by the fact that the difference between Pb isotopes of the two JENDL libraries is dominant in the comparative study, through the experimental analyses of sample reactivity by the MCNP approach. In addition, JENDL-4.0 reveals a slight difference from ENDF/B-VII.0 in all Pb isotopes and ²⁷Al, and from JEFF-3.1 in ²³⁸U and ²⁷Al. Based on these results, further experiments are needed to investigate the uncertainties of Bi isotopes with the use of the Pb–Bi and Bi plates.     

Lattice physics analysis of measured isotopic compositions of irradiated BWR 9 × 9 UO2 fuel

As part of a validation study of burnup calculations of BWR cores, lattice physics analyses were performed on burnups and isotopic compositions of U, Pu and fission product nuclides measured on five samples taken from 9 × 9 BWR fuel assemblies. Burnup calculations in infinite assembly geometry were carried out using MVP-BURN and SRAC codes coupled with major nuclear data libraries. The burnups determined based on the Nd-148 method were from 27.9 to 64.2 GWd/t. The typical relative differences in isotopic compositions (atom/Total-U) between the burnup calculations and measurements were ?2 ～ 19% for ²³⁴U, ?20 ～ 3% for ²³⁵U, ?1.5 ～ 0.1% for ²³⁶U, ?0.04 ～ 0.02% for ²³⁸U, ?4 ～ 11% for ²³⁸Pu, ?11 ～ ?2% for ²³⁹Pu, ?3 ～ 0% for ²⁴⁰Pu, ?12 ～ ?2% for ²⁴¹Pu and ?2 ～ 3% for ²⁴²Pu. They were ?2 ～ 2% for Nd isotopes, ?15 ～ 7% for Eu isotopes, ?13 ～ 1% for Cs isotopes, ?13 ～ 8% for Sm isotopes, 0 ～ 7% for ¹⁴⁷Pm, ?7 ～ ?2% for ⁹⁵Mo, ?2 ～ ?1% for¹⁰¹Ru and 0 ～ 4% for ¹⁰³Rh.     

Evaluation of neutron nuclear data on xenon isotopes

Neutron nuclear data of Xe isotopes have been evaluated in the energy region, including the resolved resonance one, from 1 keV to 20 MeV by using the theoretical nuclear reaction models. The phenomenological optical model potential was employed to calculate the total cross section for natural Xe with the coupled-channels method. The cross sections for channels of capture, (n, 2n), (n, p) and (n, α) reactions were calculated and compared with available experimental results including recently measured data. The elastic scattering angular distributions and particle emission spectra were calculated, although there is no experimental information available. Reaction cross sections of evaluated libraries were considered for comparison with the calculated results. The presently calculated cross sections reproduce better the available experimental data.     

Challenge of nuclear data for innovative nuclear energy

The article devoted to assessment of present-day demand to nuclear data for transmutation problem, including the discussion of required accuracies, status and perspectives of nuclear data evaluation and development of nuclear models. The effect of nuclear data uncertainties on radiation damage of structural materials is discussed. An analysis of ISTC projects related to nuclear data measurement and evaluation is presented. The recommendations for differential, integral experiments and recommendations on the evaluated data preparation are presented.     

核能谱数据采集接口电路

介绍一种核能谱数据采集接口，它针对核能谱数据随机变化的特点，采用多顺序程序通道去无规的方法来提高数据采集速率，在一个简单的硬件接口支持下，配置一段相应软件便可实现能谱数据的快采集。     

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.     

The F4E programme on nuclear data validation and nuclear instrumentation techniques for TBM in ITER

A dedicated effort on nuclear data validation and nuclear instrumentation techniques for Test Blanket Modules (TBM) in ITER is conducted as an integral part of Fusion for Energy's (F4E) programme to ensure validated nuclear analysis capabilities for fusion technology applications. It is closely linked to nuclear data development activities, which are jointly coordinated and conducted by F4E and nuclear data consortia formed by European research institutions. The current experimental activities, an integral copper validation experiment and gas production experiments in EUROFER elements, as well as activities on the development and testing of candidate nuclear detectors for TBM in ITER and the related design integration assessment are presented in this paper.     

Feasibility study of decay heat uncertainty reduction using nuclear data adjustment method with experimental data

The accurate prediction of the decay heat is essential, especially for nuclear power plant safety purposes. However, it is known that the decay heat predicted by nuclear fuel burn-up calculations is uncertain because of uncertainty of nuclear data employed in the calculations. If the decay heat uncertainty can be reduced, the safety margin of the predicted decay heat can also be reduced, and feasible design ranges of various types of equipments related to the decay heat can be extended. In the present study, we use the nuclear data adjustment method for the decay heat uncertainty reduction with several types of the experimental data. As a result, we clarify that the decay heat uncertainty with short- and long-term cooling periods can be reduced by this method with appropriate experimental data.     

The pioneers' legacy of inertial confinement nuclear fusion

Some years ago Professor Guillermo Velarde visited his good friend Academician Nicolai G. Basov, Nobel Prize in Physics and Director of the Lebedev Institute, and discussed the possibility of publishing a book on the History of the Inertial Confinement Nuclear fusion to be written by its Pioneers. They thought that only the own Pioneers could focus the book on a realistic and truthful manner.The years passed and unfortunately on July 1, 2001, Nicolai G. Basov passed away. In June 2002 Professor Velarde was invited to Moscow to give a Memorial lecture on Academician Basov in the Academy of Sciences. There in Moscow, Professors Vladislav Rozanov, Sergey Guskov, Natividad Carpintero-Santamaría and Guillermo Velarde decided to carry on with the project.Unfortunately, it was time to close the deadlines and the chapter of Academician Robert Dautray, former Haut-Commissaire of the CEA was not received. We regretted very much that the contribution of this very relevant Pioneer to the ICF is not included in the book. We did not receive also any material from Professor Edouard Fabre from the Ecole Polytechnique.The absence of the contributions of Nicolai G. Basov and Edward Teller was very significant. All of us would like that this ICENES 2013, hosted by the Institute of Nuclear Fusion, could be a posthumous tribute to these two outstanding Pioneers as well as a tribute to all of them.     

核动力设备可靠性数据的处理方法研究

核动力个别设备(如电动泵)在其运行历史上无失效记录或仅有很少的失效记录,仅用小样本数据反映设备可靠性参数的总体分布有所欠缺。本文提出"两步走"方法,即对核动力设备数据进行处理时参考同堆型核电站的通用数据库,将属于无信息先验的核动力设备数据处理转化为有信息先验的处理。应用Jeffreys先验模型对核动力电动泵进行贝叶斯推断,通过图检验评价模型复现观察数据的能力,结果表明模型可以完全复现观察数据具有良好的预计能力,经分析建议将第4组数据剔除后再对电动泵失效数据进行贝叶斯推断。     

Developments and needs in nuclear analysis of fusion technology

Nuclear analyses provide essential input to the conceptual design, optimisation, engineering and safety case of fusion technology in current experiments, ITER, next-step devices and power plant studies. Calculations are intricate and computer-intensive, typically requiring detailed geometry models, sophisticated acceleration algorithms, high-performance parallel computations, and coupling of large and complex transport and activation codes and databases. This paper reports progress on some key areas in the development of tools and methods to meet the specific needs of fusion nuclear analyses. In particular, advances in the production and modernisation of reference models, in the preparation and quality assurance of acceleration algorithms and coupling schemes, and in the evaluation and adaptation of alternative transport codes are presented. Emphasis is given to ITER-relevant activities, which are the main driver of advances in the field. Discussion is made of the importance of efforts in these and other areas, considering some of the more pressing needs and requirements. In some cases, they call for a more efficient and coordinated use of the scarce resources available.     

A new paradigm for core design aimed at the sustainability of nuclear energy: The solution of the extended equilibrium state

The future expansion of nuclear energy, a technology identified as one of the main candidates for reducing the world’s dependence on fossil fuels, requires a thorough analysis of the sustainability of this energy source for long-term supply. Generation-IV nuclear systems could represent a turning point for energy production by minimizing the environmental footprint of the fuel cycle. A new paradigm is thus required for reactor design, focusing, at the core design level, on both the closure of the fuel cycle and the effective utilization of natural resources.     

Uncertainty Analyses of Neutron Cross Sections for 235U in the Resonance Region

The helium embrittlement behavior of 316FR austenitic stainless steel was investigated by a tensile test at 750°C using miniature tensile specimens, which were helium-implanted below 100°C up to 5, 30, and 100 appm using a cyclotron accelerator, and were post-implantation-annealed at 750°C for 10 and 100 h. The helium-implanted specimens showed a fully intergranular fracture regardless of the helium concentration and annealing time. No microstructural changes in the as-implanted specimen up to 30 appm and formation of a small number of helium bubbles due to the post-implantation annealing were observed. The gradual release of the helium during the tensile test started after the yielding, and a sharp peak of the helium release was detected in the final fracture phase. The total number of helium atoms released was strongly dependent on the implanted helium concentration, rather than on the annealing time.     

Study of the helium-cooled lithium lead test blanket module nuclear behaviour under irradiation in ITER

The present paper deals with the detailed investigation of the helium-cooled lithium lead test blanket module (HCLL-TBM) nuclear behaviour under irradiation in ITER, carried out at the Department of Nuclear Engineering of the University of Palermo adopting a numerical approach based on the Monte Carlo method.A realistic 3D heterogeneous model of the HCLL-TBM was set-up and inserted into an ITER 3D semi-heterogeneous model that realistically simulates the reactor lay-out up to the cryostat. A Gaussian-shaped neutron source was adopted for the calculations.The main features of the HCLL-TBM nuclear response were assessed, paying a particular attention to the neutronic and photonic deposited power, the tritium production rate and the spatial distribution of their volumetric densities. Structural material irradiation damage was also investigated through the evaluation of displacement per atom and helium and hydrogen production rates.