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Tada, Kenichi; Kunieda, Satoshi; Nagaya, Yasunobu
JAEA-Data/Code 2018-014, 106 Pages, 2019/01
JAEA-Data-Code-2018-014.pdf:1.76MB
JAEA-Data-Code-2018-014-appendix(DVD-ROM).zip:6.99MB
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.
Kikuchi, Masamitsu
Hoken Butsuri, 37(1), p.12 - 14, 2002/03
no abstracts in English
*;
JNC TN8400 2001-027, 131 Pages, 2001/11
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.
Hayashida, Retsu*; Hidaka, Akihide; Nakamura, Takehiko; Kudo, Tamotsu; Otomo, Takashi; Uetsuka, Hiroshi
JAERI-Tech 2001-029, 161 Pages, 2001/03
JAERI-Tech-2001-029.pdf:9.33MB
no abstracts in English
Yoshizawa, Michio; Tsujimura, Norio*
Hoken Butsuri, 36(1), p.18 - 23, 2001/03
no abstracts in English
; *; *; *
JNC TN8410 2000-011, 185 Pages, 2000/05
JNC-TN8410-2000-011.pdf:4.67MB
This report describes the neutronic design calculational methods used in Fuel Design and Evaluation Group in order to inform other related sections of FBR core analysis technology and hand down the technology. Especially we show the neutronics calculation procedures used for the conceptual design study of the advanced core with 127 pin bundle for MONJU that has been carried out in our group. The topics include effective cross section preparation calculations, two-dimensional depletion calculations, three-dimensional diffusion calculations, reactivity coefficient calculations, and control rod worth calculations. The calculational methods shown in this report are the standard neutronics calculation methods employed in our group at the moment. However, the improvement of calculation codes, the reduction of correction factors and uncertainties for design using the nuclear data obtained in the start-up test for MONJU and so on, and the update of nuclear data file will be planned in order to improve evaluation accuracies. Those may change the neutronic design calculational methods, but we decided to describe the present standard calculational methods in our group from the viewpoint of sharing information in JNC.
Tanabe, Fumiya
SCIaS, 5(3), p.22 - 23, 2000/03
no abstracts in English
Ohno, Shuji; Matsuki, Takuo*; ; Miyake, Osamu
JNC TN9520 2000-001, 196 Pages, 2000/01
JNC-TN9520-2000-001.pdf:5.13MB
ASSCOPS (Analysis of Simultaneous Sodium Combustion in Pool and Spray) has been developed for analyses of thermal consequences of sodium leak and fire accidents in LMFBRs. This report presents a description of the computational models, input and output data as the user's manual of ASSCOPS version 2.1. ASSCOPS is an integrated computational code based on the sodium pool fire code SOFIRE II developed by the Atomics International Division of Rockwell International, and on the sodium spray fire code SPRAY developed by the Hanford Engineering Development Laboratory in the U.S. The users of ASSCOPS need to specify the sodium leak conditions (leak flow rate and temperature, etc.), the cell geometries (cell volume, surface area and thickness of structures, etc.), and the atmospheric initial conditions such as gas temperature, pressure, and composition. ASSCOPS calculates the time histories of atmospheric temperature, pressure and of structural temperature.
Takahashi, Tomoyuki
JAERI-Data/Code 98-003, 74 Pages, 1998/02
JAERI-Data-Code-98-003.pdf:2.79MB
no abstracts in English
; Ohno, Shuji; Miyake, Osamu; ; Seino, Hiroshi
PNC TN9520 97-001, 185 Pages, 1997/12
ASSCOPS(Analysis of Simultaneous Sodium Combustion in Pool and Spray) has been developed for analyses of thermal consequences of sodium leak and fire accidents in LMFBRs. This report presents a description of the computational models, input, and output as the user's manual of ASSCOPS version 2.0. ASSCOPS is an integrated code based on the sodium pool fire code SOFIRE II developed by the Atomics International Division of Rockwell International, and the sodium spray fire code SPRAY developed by the Hanford Engineering Development Laboratoly in the U.S. The experimental studies conducted at PNC have been reflected in the ASSCOPS improvement. The users of ASSCOPS need to specify the sodium leak conditions (leak flow rate and temperature, etc.), the cell geometries (volume and structure surface area and thickness, etc.), and the atmospheric initial conditions, such as gas temperature, pressure, and gas composition. ASSCOPS calculates the time histories of atmospheric pressure and temperature changes along with those of the structural temperatures.
Ando, Hideki;
PNC TN9100 96-006, 23 Pages, 1996/02
None
*; Aoyagi, Tetsuo; *; *; Tani, Keiji
JAERI-M 94-040, 90 Pages, 1994/03
no abstracts in English
; ; ; ; ; ;
PNC TN8520 93-003, 410 Pages, 1994/01
This manual includes the standard procedures for analysis of radioactive materials and chemical polluants in liquid and gaseous waste elluent discharged into the environment from the nuclear facilities of Tokai Works of PNC. The third edition, PNC N852-84-06 was published in 1984. Almost all analytical procedures have been modified and new technique has been used for nine years, so the forth edition revised entirely was to be published this time. And most of analytical procedures will be revised with the point of improving of reliabiliy.
