An Estimation method for an unknown covariance in cross-section adjustment based on unbiased and consistent estimator

Maruyama, Shuhei; Endo, Tomohiro*; Yamamoto, Akio*

Journal of Nuclear Science and Technology, 60(11), p.1372 - 1385, 2023/11

https://doi.org/10.1080/00223131.2023.2203707

Great achievements of M. Salvatores for nuclear data adjustment study with use of integral experiments

Yokoyama, Kenji; Ishikawa, Makoto*

Annals of Nuclear Energy, 154, p.108100_1 - 108100_11, 2021/05

https://doi.org/10.1016/j.anucene.2020.108100

In the design of innovative nuclear reactors such as fast reactors, the improvement of the prediction accuracies for neutronics properties is an important task. The nuclear data adjustment is a promising methodology for this issue. The idea of the nuclear data adjustment was first proposed in 1964. Toward its practical application, however, a great deal of study has been conducted over a long time. While it took about 10 years to establish the theoretical formulation, the research and development for its practical application has been conducted for more than half a century. Researches in this field are still active, and the fact suggests that the improvement of the prediction accuracies is indispensable for the development of new types of nuclear reactors. Massimo Salvatores, who passed away in March 2020, was one of the first proposers to develop the nuclear data adjustment technique, as well as one of the great contributors to its practical application. Reviewing his long-time works in this area is almost the same as reviewing the history of the nuclear data adjustment methodology. The authors intend that this review would suggest what should be done in the future toward the next development in this area. The present review consists of two parts: a) the establishment of the nuclear data adjustment methodology and b) the achievements related to practical applications. Furthermore, the former is divided into two aspects: the study on the nuclear data adjustment theory and the numerical solution for sensitivity coefficient that is requisite for the nuclear data adjustment. The latter is separated to three categories: the use of integral experimental data, the uncertainty quantification and design target accuracy evaluation, and the promotion of nuclear data covariance development.

Comparative study on prediction accuracy improvement methods with the use of integral experiments for neutronic characteristics of fast reactors

Yokoyama, Kenji; Kitada, Takanori*

Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.1221 - 1230, 2018/04

no abstracts in English

Methods and approaches to provide feedback from nuclear and covariance data adjustment for improvement of nuclear data files

Palmiotti, G.*; Salvatores, M.*; Yokoyama, Kenji; Ishikawa, Makoto

NEA/NSC/R(2016)6 (Internet), 42 Pages, 2017/05

An Integrated approach to source term uncertainty and sensitivity analysis for nuclear reactor severe accidents

Zheng, X.; Ito, Hiroto; Tamaki, Hitoshi; Maruyama, Yu

Journal of Nuclear Science and Technology, 53(3), p.333 - 344, 2016/03

AA2014-0796.pdf:0.84MB

https://doi.org/10.1080/00223131.2015.1044262

Status of the JENDL project

Shibata, Keiichi; Nakagawa, Tsuneo; Fukahori, Tokio; Ichihara, Akira; Iwamoto, Osamu; Otsuka, Naohiko*; Katakura, Junichi

AIP Conference Proceedings 769, p.171 - 176, 2005/05

https://doi.org/10.1063/1.1944983

The presentation deals with the activities on JENDL. JENDL-4 is being developed as a general purpose library. This library will include charged-particle and photon induced reaction data as well as spontaneous fission data for a limited number of nuclei in addition to neutron induced reaction data. The maximum incident energy can be extended to higher values than 20 MeV depending on data needs. Key subjects are improvements of accuracy of minor actinide and FP data, evaluations of covariances and so on. Quality assurance is regarded as important for JENDL-4. As special purpose files, we released the JENDL High Energy File and Photonuclear Data File this year. For the development of ADS, the Actinide File is being made. Furthermore,covariance data for some nuclei are also evaluated for ADS. Nuclear model codes are developed in order to reflect recent advances in nuclear theory on data evalulations. For users' covenience, we are working on the development of the Combined System for Nuclear Data Utilization, Circulaiton and Transfer.

Evaluations of heavy nuclide data for JENDL-3.3

Kawano, Toshihiko*; Matsunobu, Hiroyuki*; Murata, Toru*; Zukeran, Atsushi*; Nakajima, Yutaka*; Kawai, Masayoshi*; Iwamoto, Osamu; Shibata, Keiichi; Nakagawa, Tsuneo; Osawa, Takaaki*; et al.

JAERI-Research 2003-026, 53 Pages, 2003/12

JAERI-Research-2003-026.pdf:2.48MB

New evaluations of neutron nuclear data for Uranium, Plutonium, and Thorium isotopes which are essential for applications to nuclear technology were carried out for the Japanese Evaluated Nuclear Data Library, JENDL-3.3. The objectives of the current release of JENDL were to fix several problems which have been reported for the previous version, to improve the accuracy of the data, and to evaluate covariances for the important nuclides. Quantities in JENDL-3.2 were extensively re-evaluated or replaced by more reliable values. The heavy nuclide data in JENDL-3.3 were validated with several benchmark tests, and it was reported that the current release gave a good prediction of criticalities.

3-D shielding calculation method for 1-MW JSNS

Maekawa, Fujio; Tamura, Masaya

Proceedings of ICANS-XVI, Volume 3, p.1051 - 1058, 2003/07

A three-dimensional (3-D) shielding calculation model for MCNPX was produced for shielding design of 1-MW JSNS. The model included simplified target-moderator-reflector assembly, helium-vessel and neutron beam extraction pipes, shutters, shield blocks, gaps and void spaces between these components, and so on, and could treat streaming effects precisely. The particle splitting and kill method with cell importance parameters was adopted as a variance reduction method. The cell importance parameters for such a large target station of about 15 m in diameter and 12 m in hight in which neutron fluxes attenuated more than 12 orders of magnitude could be determined appropriately by automated iteration calculations. This calculation procedure enabled detailed 3-D shielding design calculations for the whole target station in a short time, i.e., within 2 days, and contributed for progress of shielding designs of JSNS.

Evaluation of shutdown -ray dose rates around the duct penetration by three-dimensional Monte Carlo decay -ray transport calculation with variance reduction method

Sato, Satoshi; Iida, Hiromasa; Nishitani, Takeo

Journal of Nuclear Science and Technology, 39(11), p.1237 - 1246, 2002/11

https://doi.org/10.3327/jnst.39.1237

no abstracts in English

JENDL Dosimetry File 99 (JENDL/D-99)

Kobayashi, Katsuhei*; Iguchi, Tetsuo*; Iwasaki, Shin*; Aoyama, Takafumi*; Shimakawa, Satoshi; Ikeda, Yujiro; Odano, Naoteru; Sakurai, Kiyoshi; Shibata, Keiichi; Nakagawa, Tsuneo; et al.

JAERI 1344, 133 Pages, 2002/01

JAERI-1344.pdf:7.59MB

The JENDL Dosimetry File 99 (JENDL/D-99) has been prepared for determinations of neutron flux/fluence and energy spectrum at specific neutron fields. This file contains data for 67 reactions with 47 nuclides. Cross sections for 33 major dosimetry reactions and their covariance data were simultaneously generated and the other 34 reaction data were mainly adopted from the first version, JENDL/D-91. The GMA code was mainly used for most of the evaluation procedures by referring the basic experimental data in EXFOR. The resultant data are given in the neutron energy region below 20 MeV in both of point-wise and group-wise files in the ENDF-6 format. In order to confirm reliability of the data, several integral tests have been carried out: comparison with the data in IRDF-90V2 and average cross sections measured in fission neutron fields, fast/thermal reactor spectra, DT neutron fields and Li(d,n) neutron fields. The contents of JENDL/D-99 and the results of the integral tests are described in this report. All of the dosimetry cross sections are shown in a graphical form in the Appendix.