JAEA-Technology 2021-023, 190 Pages, 2021/11
Computational analyses on nuclear criticality characteristics were carried out for heterogeneous lattice systems composed of water moderator and fuel rods utilized in lowpower research and test reactors, in which the depletion of fuel due to burnup is relatively small, by using the continuous-energy Monte Carlo code MVP Version 2 with the evaluated nuclear data library JENDL-4.0. In the analyses, the minimum critical number of fuel rods was evaluated using calculated neutron multiplication factors for the heterogeneous systems of the uranium dioxide fuel rod in the Static Experiments Critical Facility (STACY) and the Tank-type Critical Assembly (TCA), and the uranium-zirconium hydride fuel rod in the Nuclear Safety Research Reactor (NSRR). In addition, six sorts of the ratio of reaction rates, which are components of neutron multiplication factors, were calculated in the analyses to explain the variation of neutron multiplication factors with the ratio of water moderator to fuel volume in a unit fuel rod cell. Those results of analyses are considered to be useful for the confirmation of reasonableness and validity of criticality safety measures as data showing criticality characteristics for water-moderated heterogeneous lattice systems composed of the existing fuel rods in research and test reactors, of which criticality data are not sufficiently provided by the Criticality Safety Handbook.
Kaku Deta Nyusu (Internet), (130), p.29 - 34, 2021/10
This manuscript describers the appearance of Japanese Evaluated Nuclear Data Library (JENDL) for Europe, the status of the main nuclear data library of European countries, i.e., Joint Evaluated Fission and Fusion (JEFF) Nuclear Data Library and the future of evaluation of the nuclear data, based on the experience of working at OECD/NEA Data Bank which manages the development of JEFF.
Takeshita, Hayato; Meigo, Shinichiro; Matsuda, Hiroki; Iwamoto, Hiroki; Maekawa, Fujio; Watanabe, Yukinobu*
JPS Conference Proceedings (Internet), 33, p.011045_1 - 011045_6, 2021/03
To improve accuracy of nuclear design of accelerator driven nuclear transmutation systems, nuclide production cross sections on Ni and Zr, which were candidate materials to be used in ADS, were measured for GeV energy protons. The measured results were compared with PHITS calculations and JENDL/HE-2007.
Matsuda, Hiroki; Takeshita, Hayato*; Meigo, Shinichiro; Maekawa, Fujio; Iwamoto, Hiroki
JPS Conference Proceedings (Internet), 33, p.011047_1 - 011047_6, 2021/03
Accurate nuclide production cross-section data are required for the design of Accelerator-Driven nuclear transmutation System (ADS) such as the design of radioactive waste disposal, design of remote-handling procedure of highly activated components, and evaluation of exposure doses of rad-workers. Although much efforts have been devoted to nuclide production cross-section measurements so far, uncertainties of the measured data are sometimes large as several tens percentage, and there is no experimental data in the GeV energy region even for some of important nuclides. In this study, proton induced nuclide production cross-section of iron, which is the most important constituent element of steel, was measured. The present experiment was compared with calculations by the PHITS code with several physics models including Bertini and INCL4.6 and evaluated nuclear data JENDL-HE/2007. The most significant discrepancy found in this study was the production cross sections via the (p,xn) reaction. It was suggested that further improvements, such as the in-medium effect on the nucleon-nucleon scattering and the Pauli blocking, were required in the intra-nuclear cascade models used in this study.
JAEA-Conf 2020-001, p.11 - 16, 2020/12
Matsuda, Hiroki; Meigo, Shinichiro; Iwamoto, Hiroki; Maekawa, Fujio
EPJ Web of Conferences, 239, p.06004_1 - 06004_4, 2020/09
For the Accelerator-Driven nuclear transmutation System (ADS), nuclide production yield estimation in the lead-bismuth target is important to manage the target. However, experimental data of nuclide production yield by spallation and high-energy fission reactions are scarce. In order to obtain the experimental data, we experimented in J-PARC using Pb and Bi samples. The samples were irradiated with protons at various kinematic energy points between 0.4 and 3.0 GeV. After the irradiation, the nuclide production cross section over Be to Re was obtained by spectroscopic measurement of decay gamma-rays from the samples with HPGe detectors. The present experimental results were compared with the evaluated data (JENDL-HE/2007) and the calculation with the PHITS code and the INCL++ code. The present experiment data showed consistency with other experimental data with better accuracy than other ones. In reactions to produce light nuclides, JENDL and calculation with the PHITS and INCL++ for Be production agreed with the data.Na production, however, underestimated about 1/10 times. For middle to heavy nuclide productions cases, both calculations agreed with the experiment by a factor of two. JENDL showed lower energy having a maximum value of excitation function maximal value than the experimental data.
Sugino, Kazuteru; Numata, Kazuyuki*; Ishikawa, Makoto; Takeda, Toshikazu*
Annals of Nuclear Energy, 130, p.118 - 123, 2019/08
In MA sample irradiation test data calculations, the neutron fluence during irradiation period is generally scaled by using dosimetry data in order to improve calculation accuracy. In such a case, appropriate correction is required to burnup sensitivity coefficients obtained by the conventional generalized perturbation theory because some cancellations occur in the burnup sensitivity coefficients. Therefore, a new formula for the burnup sensitivity coefficient has been derived with the consideration of the neutron fluence scaling effect (NFS). In addition, the cross-section-induced uncertainty is evaluated by using the obtained burnup sensitivity coefficients and the covariance data based on the JENDL-4.0.
Yokoyama, Kenji; Sugino, Kazuteru; Ishikawa, Makoto; Maruyama, Shuhei; Nagaya, Yasunobu; Numata, Kazuyuki*; Jin, Tomoyuki*
JAEA-Research 2018-011, 556 Pages, 2019/03
We have developed a new unified cross-section set ADJ2017, which is an improved version of the unified cross-section set ADJ2010 for fast reactors. The unified cross-section set is used for reflecting information of C/E values (analysis / experiment values) obtained by integral experiment analyses; the values are stored in the standard database for FBR core design via the cross-section adjustment methodology, which integrates with the information such as uncertainty (covariance) of nuclear data, uncertainty of integral experiment / analysis, sensitivity of integral experiment with respect to nuclear data. The ADJ2017 is based on Japan's latest nuclear data library JENDL-4.0 as in the previous version of ADJ2010, and it incorporates more information on integral experimental data sets related to minor actinides (MAs) and degraded plutonium (Pu). In the creation of ADJ2010, a total of 643 integral experimental data sets were analyzed and evaluated, and 488 of the integral experimental data sets were finally selected to be used for the cross-section adjustment. In contrast, we have evaluated a total of 719 data sets, and eventually adopted 620 integral experimental data sets to create ADJ2017. ADJ2017 shows almost the same performance as ADJ2010 for the main neutronic characteristics of conventional sodium-cooled MOX-fuel fast reactors. In addition, for the neutronic characteristics related to MA and degraded Pu, ADJ2017 improves the C/E values of the integral experimental data sets, and reduces the uncertainty induced by the nuclear data. ADJ2017 is expected to be widely used in the analysis and design research of fast reactors. Moreover, it is expected that the integral experimental data sets used for ADJ2017 can be utilized as a standard database of FBR core design.
Tada, Kenichi; Kunieda, Satoshi; Nagaya, Yasunobu
JAEA-Data/Code 2018-014, 106 Pages, 2019/01
A new nuclear data processing code FRENDY has been developed in order to process the evaluated nuclear data library JENDL. Development of FRENDY helps to disseminate JENDL and various nuclear calculation codes. FRENDY is developed not only to process the evaluated nuclear data file but also to implement the FRENDY functions to other calculation codes. Users can easily use many functions e.g., read, write, and process the evaluated nuclear data file, in their own codes when they implement the classes of FRENDY to their codes. FRENDY is coded with considering maintainability, modularity, portability and flexibility. The processing method of FRENDY is similar to that of NJOY. The current version of FRENDY treats the ENDF-6 format and generates the ACE file which is used for the continuous energy Monte Carlo codes such as PHITS and MCNP. This report describes the nuclear data processing methods and input instructions for FRENDY.
JAEA-Conf 2018-001, p.87 - 91, 2018/12
Status and plan of JENDL will be presented. After the release of JENDL-4.0 in 2010, six special purpose files have been developed. Four of them were already released and two are under preparation for the release. New decay and yield data for fission products were released as JENDL/FPD-2011 and JENDL/FPY-2011 in 2011, respectively. JENDL-4.0/HE released in 2015 includes proton and neutron induced reaction data up to 200 MeV. Comprehensive decay data were released as JENDL/DDF-2015 which contains data for 3,237 nuclides. New photonuclear reaction data JENDL/PD-2016 and an activation file JENDL/AD-2017 are under preparation for release. Regarding general purpose file, two activities are in progress. One is JENL-4.0u which is created for maintenance of JENDL-4.0 and the other is development of next version of JENDL. For the next JENDL, evaluation for light nuclei and structure material are in progress. It is planed that next version of JENDL will be JENDL-5 which contains nuclear data for all nuclei having natural abundance. Addition of covariance data will be one of the main targets.
Iwamoto, Hiroki; Stankovskiy, A.*; Fiorito, L.*; Van den Eynde, G.*
European Physical Journal; Nuclear Sciences & Technologies (Internet), 4, p.42_1 - 42_7, 2018/11
This paper presents a nuclear data sensitivity and uncertainty analysis of the effective delayed neutron fraction for critical and subcritical cores of the MYRRHA reactor using the continuous-energy Monte Carlo transport code MCNP. The sensitivities are calculated by the modified -ratio method proposed by Chiba. Comparing the sensitivities obtained with different scaling factors introduced by Chiba shows that a value of is the most suitable for the uncertainty quantification of . Using the calculated sensitivities and the JENDL-4.0u covariance data, the uncertainties for the critical and subcritical cores are determined to be 2.2 0.2% and 2.0 0.2%, respectively, which are dominated by delayed neutron yield of Pu and U.
Matsuda, Hiroki; Meigo, Shinichiro; Iwamoto, Hiroki
Journal of Nuclear Science and Technology, 55(8), p.955 - 961, 2018/08
We have started an experimental program to measure activation cross sections systematically in the proton-induced spallation reaction in structural materials commonly used in high-intensity proton accelerator-based facilities, such as Japan Proton Accelerator Research Complex (J-PARC). As the first step of the program, aluminum (Al) was chosen to verify the adequacy of the measurement technique implemented in a J-PARC proton beam environment because data of Al have been relatively well studied both by experimental measurement and simulation. Activation cross sections of Be, Na, and Na in Al were measured at proton energy points from 0.4, 1.3, 2.2 to 3.0 GeV, which could be delivered smoothly from the synchrotron. The validity of experimental data has been verified by introducing an effective proton numbers determination procedure. We compared the measured data with existing experimental data, the evaluated data (JENDL-HE/2007), and the calculations with several intra-nuclear cascade models by the Particle and Heavy Ion Transport code System (PHITS) code. Although the experimental data agreed with JENDL-HE/2007, the calculations underestimated about 40%. This could come from the evaporation model (generalized evaporation model) being implemented in the PHITS code. We found that the calculations agreed with the experimental data by an upgraded PHITS code.
Usami, Shin; Kishimoto, Yasufumi*; Taninaka, Hiroshi; Maeda, Shigetaka
JAEA-Technology 2018-003, 97 Pages, 2018/07
The decay heat used for effectiveness evaluation of the prevention measures against severe accidents in the prototype fast breeder reactor Monju was evaluated by applying the updated nuclear data libraries based on JENDL-4.0, reflecting the realistic core operation pattern, and setting the rational extent of uncertainty. The decay heats of fission products, the actinide nuclides such as Cm-242, and radioactive structural materials were calculated by FPGS code. The decay heat of U-239 and Np-239 was evaluated based on ANSI/ANS-5.1-1994. The calculation uncertainty of each decay heat was evaluated based on summation of uncertainty factors, C/E values of reaction rates obtained in Monju system startup test, and so on. Furthermore, the decay heat evaluation method based on the FPGS90 was verified by the comparison of the results of the decay heat measurement of the two spent MOX fuel subassemblies in the experimental fast reactor Joyo MK-II core.
Journal of Nuclear Science and Technology, 55(6), p.614 - 622, 2018/06
Toward the development of the next version of Japanese Evaluated Nuclear Data Library (JENDL) general-purpose file, we calculate neutron cross-sections on Cu from 50 keV to 20MeV, which is the incident energy range above the resolved resonance region in JENDL-4.0. A dispersive optical model potential is adopted with a coupled-channel method for interaction between neutron and Cu. Direct, pre-equilibrium, and compound processes are taken into account in the calculation. All cross-sections, differential and double-differential cross-sections are consistently calculated with a single set of model parameters. The calculation results reproduce the measured data very well. In addition, disagreement between the calculated and experimental values seen in an integral test for the Cu reaction is improved by using the cross-section data obtained from the present work.
Meigo, Shinichiro; Matsuda, Hiroki; Iwamoto, Hiroki
Proceedings of 13th International Topical Meeting on Nuclear Applications of Accelerators (AccApp '17) (Internet), p.396 - 402, 2018/05
no abstracts in English
Chadwick, M. B.*; Capote, R.*; Trkov, A.*; Herman, M. W.*; Brown, D. A.*; Hale, G. M.*; Kahler, A. C.*; Talou, P.*; Plompen, A. J.*; Schillebeeckx, P.*; et al.
Nuclear Data Sheets, 148, p.189 - 213, 2018/02
The CIELO collaboration has studied neutron cross sections on nuclides that significantly impact criticality in nuclear facilities - U, U, Pu, Fe, O and H - with the aim of improving the accuracy of the data and resolving previous discrepancies in our understanding. This multi-laboratory pilot project, coordinated via the OECD/NEA Working Party on Evaluation Cooperation (WPEC) Subgroup 40 with support also from the IAEA, has motivated experimental and theoretical work and led to suites of new evaluated libraries that accurately reflect measured data and also perform well in integral simulations of criticality. This report summarizes our results and outlines plans for the next phase of this collaboration.
JAEA-Conf 2017-001, p.103 - 108, 2018/01
In the Nuclear Data Center of Japan Atomic Energy Agency (JAEA-NDC), we are engaged in the evaluation activity for the next version of the Japanese Evaluated Nuclear Data Library, JENDL-4.0. Zirconium is an important structural material in nuclear reactors, and zircaloys are being employed in fuel rods. Also, Zr is a long-lived fission product (LLFP) with a half-life of 1.6110 years. At present we are investigating resonance parameters of Zr isotopes using experimental data published after the evaluation of JENDL-4.0. Through this work, a negative resonance of Zr in JENDL-4.0 was removed to reproduce the J-PARC/MLF/ANRRI experiment. The resonance parameters for other natural Zr isotopes will be altered by adopting the data obtained at CERN n-TOF experiments.