Another important piece; One point burnup calculation code as a Killer Application

Suyama, Kenya; Yokoyama, Kenji

Kaku Deta Nyusu (Internet), (119), p.38 - 47, 2018/02

http://www.aesj.or.jp/~ndd/ndnews/pdf119/No119-05.pdf

We have developed numerous neutronics calculation codes in Japan. However, development of the one-point burnup calculation code which replaces the still widely used ORIGEN2 code has not been successful. The one point burnup code is indispensable to evaluate the characteristics of the used nuclear fuel increasing in Japan, and it uses all evaluated nuclear data including the fission yield and decay data as well as cross section data. It means that it could be the Killer Application in the field of the nuclear data and neutronics code. This report describes the necessity of the one point burnup calculation code development in Japan and required function and performance which have been considered by authors.

Theoretical model analysis of reactions on Be and C at incident energies up to 50 MeV

Nakayama, Shinsuke; Kono, Hiroshi*; Watanabe, Yukinobu*; Iwamoto, Osamu; Ogata, Kazuyuki*

Physical Review C, 94(1), p.014618_1 - 014618_9, 2016/07

https://doi.org/10.1103/PhysRevC.94.014618

Double-differential thick target neutron yields (TTNYs) from deuteron bombardment on thick Be and C targets are analyzed using the DEURACS (DEUteron-induced Reaction Analysis Code System). The calculated TTNYs reproduced the experimental ones quantitatively well in the incident energy range up to 50 MeV. In addition, it was found that the proton stripping reaction makes the most dominant contribution to neutron production. From the analysis, we conclude that the DEURACS is applicable to reactions and modeling of the stripping reaction is essential to predict neutron production yields accurately.

Review of simulation codes applicable to evaluation of high energy neutron dose

Sato, Tatsuhiko

Hoshasen, 31(4), p.313 - 318, 2005/10

http://ci.nii.ac.jp/naid/10016952728

Evaluation of high-energy neutron dose is one of the key issues in the shielding design of accelerator facilities and in the planning of long-term space missions. High-energy neutron transport simulation codes play an important role in the evaluation, since there is a large difficulty in the precise measurement of high-energy neutron doses. This paper reviews the Monte-Carlo simulation codes applicable to the purpose, and summarizes the requirements for the future development of the codes.

Summaries of research and development activities by using JAERI computer system in FY2003 (April 1, 2003 - March 31, 2004)

Center for Promotion of Computational Science and Engineering

JAERI-Review 2005-008, 199 Pages, 2005/03

JAERI-Review-2005-008.pdf:30.15MB

The Center for Promotion of Computational Science and Engineering (CCSE) in Japan Atomic Energy Research Institute (JAERI) installed large computer system including super-computers in order to support research and development activities in JAERI. CCSE is also operate anad manage the computer system and network system. This report presents the survey results of research and development activities by using JAERI computer system in FY2003.

Equipment of model template for a code predicting the migration of ground additions, MOGRA

Takahashi, Tomoyuki*; Amano, Hikaru; Uchida, Shigeo*; Ikeda, Hiroshi*; Matsuoka, Shungo*; Hayashi, Hiroko*; Kurosawa, Naohiro*

Kankyo Eisei Kogaku Kenkyu, 17(3), p.340 - 344, 2003/07

no abstracts in English

Multi-group diffusion perturbation calculation code PERKY; 2002

Iijima, Susumu; Okajima, Shigeaki

JAERI-Data/Code 2002-023, 44 Pages, 2002/12

JAERI-Data-Code-2002-023.pdf:1.6MB

no abstracts in English

Theoretical background and user's manual for the computer code on groundwater flow and radionuclide transport calculation in porous rock

*;

JNC TN8400 2001-027, 131 Pages, 2001/11

JNC-TN8400-2001-027.pdf:0.8MB

In order to document a basic manual about input data, output data, execution of computer code on groundwater flow and radionuclide transport calculation in heterogeneous porous rock, we investigated the theoretical background about geostastical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport which calculates water flow in three dimension, the path of moving radionuclide, and one dimensional radionuclide migration. In this report, based on above investigation we describe the geostastical background about simulating heterogeneous permeability field. And we describe construction of files, input and output data, a example of calculating of the programs which simulates heterogeneous permeability field, and calculates groundwater flow and radionuclide transport. Therefore, we can document a manual by investigating the theoretical background about geostastical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport calculation. And we can model heterogeneous porous rock and analyze groundwater flow and radionuclide transport by utilizing the information from this report.

Technical development on burn-up credit for spent LWR fuels

Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori

JAERI-Tech 2000-071, 381 Pages, 2000/10

JAERI-Tech-2000-071.pdf:17.6MB

no abstracts in English

Expansion of material balance analysis function on nuclear fuel cycle

Ohtaki, Akira; ; ; *; *;

JNC TN9410 2000-006, 74 Pages, 2000/04

JNC-TN9410-2000-006.pdf:3.01MB

To evaluate materials balance in nuclear fuel cycle quickly and quantitatively the fuel cycle requirement code "FAMILY" was improved. And an accumulated TRU&LLFP quantity analysis code was developed. The contents are as follows: (1)A calculation ability of minor actinide production and expenditure was added to the "FAMILY" code. (2)An output program for the "FAMILY" calculation results was developed. (3)A simple version of "FAMILY" code was developed. (4)An analysis code for accumulated TRU&LLFP quantity in nuclear fuel cycle was developed.

ComparaUve analyses on nuclear charaderistics of water-cooled breeder cores

; Sato, Wakaei*;

JNC TN9400 2000-037, 87 Pages, 2000/03

JNC-TN9400-2000-037.pdf:3.48MB

ln order to compare the nuclear characteristics of water-cooled bleeder cores with that of LMFBR, MOX fuel cell models are established for boiling and non-boiling LWR, non-boiling HWR and sodium-cooled reactor. Frst, the comarison is made between the heterogeneous cell calculation results by SRAC and those by SLAROM. The results show some differences as for neutron energy spectrum, one-grouped cross section and conversion ratio due to the different grouped cross section library (both are based on JENDL-3.2, though) used for each code, however, the difference is acceptably small for grasping the basic characteristics of the above-mentioned cores. Second, using the SLAROM code, main core parameters such as mean neutron energy, ratio of fast neutron and -value, are analyzed. The comparison between the cores show that softened neutron spectrum by the scattering effect of hydrogen or heavy hydrogen increase the contribution of nuclear reaction (especially for neutron capture reaction rather than fission reaction) in lower energy region comparing with LMFBR. ln order to overcome the effect, tighter lattice than LMFBR is necessary for water-cooled cores to realize the breeding of fissile nuclides. Third, effects of Pu isotopic composition on the breeding ratio are evaluated using SRAC burnup calculation. From the results, it is confirmed that degraded Pu (larger ratio of Pu-240) show the larger breeding ratio. At last, sensitivity analyses are made for k-effective and main reaction ratios. As for k-effective, using a temporary covariance data of JENDL-3.2, uncertainty resulting from the cross sections' error is analyzed for a boiling LWR and a sodium-cooled reactor. The boiling LWR core shows larger sensitivity in lower energy region than the sodium-cooled reactor (especially for the energy region lower than 1kev), And, 18-group analysis that is considered acceptably good for LMFBR analysis, should not be enough for accurate sensitivity estimation of ...