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Nakagawa, Tsuneo
JAERI-Data/Code 2003-016, 89 Pages, 2003/11
JAERI-Data-Code-2003-016.pdf:3.0MB
The data storage and retrieval system for neutron-induced reaction data NESTOR2 was developed. NESTOR2 was written in Fortran77, and can be used on workstations with UNIX and personal computers with Windows. Input data to NESTOR2 are those in EXFOR which are compiled and maintained under the international collaboration for nuclear data. By using the retrieval code, they can obtain lists of data index, numerical data and comment information and data files of numerical data. The present report explains the system and provides a user's manual of NESTOR2.
Brown, B.*; Nagaya, Yasunobu
Transactions of the American Nuclear Society, 87, p.230 - 232, 2002/11
no abstracts in English
Sato, Haruo
JNC TN8410 2001-003, 40 Pages, 2001/01
JNC-TN8410-2001-003.pdf:1.13MB
A program (TDROCK1.FOR) for simulation and analysis of through-diffusion experiments for a single layer of diffusion media was developed. This program was made by Pro-Fortran language, which was suitable for scientific and technical calculations, and relatively easy explicit difference method was adopted for an analysis. In the analysis, solute concentration in the tracer cell as a function of time that we could not treat to date can be input and the decrease in the solute concentration as a function of time by diffusion from the tracer cell to the measurement cell, the solute concentration distribution in the porewater of diffusion media and the solute concentration in the measurement cell as a function of time can be calculated. In addition, solution volume in both cells and diameter and thickness of the diffusion media are also variable as an input condition. This simulation program could well explain measured result by simulating solute concentration in the measurement cell as a function of time for case which apparent and effective diffusion coefficients were already known. Based on this, the availability and applicability of this program to actual analysis and simulation were confirmed. This report describes the theoretical treatment for the through-diffusion experiments for a single layer of diffusion media, analytical model, an example of source program and the manual.
Satmoko, A.*;
JNC TN9400 99-035, 37 Pages, 1999/04
In sodium environment, materia1 316FR stainless steel risks to suffer from carburization. In this study, an analysis using a Fortran program is conducted to evaluate the carbon influence on the creep behavior of 316FR based on experimental results from uni-axial creep test that had been performed at temperature 550
C in sodium environment simulating Fast Breeder Reactor condition. As performed in experiments, two parts are distinguished. At first, elastic-plastic behavior is used to simulate the fact that just before the beginning of creep test, specimen suffers from load or stress much higher than initial yield stress. In second part, creep condition occurs in which the applied load is kept constant. The plastic component should be included, since stresses increase due to section area reduction. For this reason, elastic-plastic-creep behavior is considered. Through time carbon penetration occurs and its concentration is evaluated empirically. This carburization phenomena are assumed to affect in increasing yield stress, decreasing creep strain rate, and increasing creep rupture strength of material. The model is capable of simulating creep test in sodium environment. Material near from surface risks to be carburized. Its material properties change leading to non-uniform distribution of stresses. Those layers of material suffer from stress concentration, and are subject to damage. By introducing a damage criteria, crack initialization can thus be predicted. And even, crack growth can be evaluated. For high stress levels, tensile strength criterion is more important than creep damage criterion. But in low stress levels, the latter gives more influence in fracture. Under high stress, time to rupture of a specimen in sodium environment is shorter than in air. But for stresses lower than 26 kgfmm
, the time to rupture of creep in sodium environment is the same or little longer than in air. Quantitatively, the carburization effect at ...
*; *; *; *
JAERI-Data/Code 99-014, 47 Pages, 1999/03
JAERI-Data-Code-99-014.pdf:2.61MB
no abstracts in English
Yamane, Tsuyoshi; *
JAERI-Data/Code 99-011, 46 Pages, 1999/03
JAERI-Data-Code-99-011.pdf:1.33MB
no abstracts in English
Watanabe, Tadashi; Ebihara, Kenishi; Kaburaki, Hideo
JAERI-Data/Code 97-056, 32 Pages, 1998/01
JAERI-Data-Code-97-056.pdf:1.57MB
no abstracts in English
*
JAERI-Data/Code 96-006, 49 Pages, 1996/03
JAERI-Data-Code-96-006.pdf:1.51MB
no abstracts in English
gas turbine power generation system (User's manual); Sekiguchi, Nobutada
PNC TN9520 95-002, 66 Pages, 1995/02
This analysis program code STEDFAST; Space, TErrestrial and Deep sea FAST reactor gas tubine system; is used to get the adequate values of system parameters on fast reactor gas turbine power generation systems used as power sources for deep sea, space and terrestrial cogeneration. Characteristics of the code are as follows. Objective systems of the code are a deep sea, a space and a terrestrial reactors. Primary coolants of the systems are NaK, Na, Pb and Li. Secondary coolant is the mixture gas of He and Xe. The ratio of He and Xe is arbitrary. Modeling of components in the systems was performed so that detailed modeling might be capable in future and that a transient analytical code could be easily made by using the code. A progra㎜ing language is MAC-FORTRAN. The code can be easily used in a personal computer. The code made possible instant calculation of various state values in a Brayton cycle, understanding the effects of many parameters on thermal efficiency and finding the most adequate values of the parameters. From now on, detailed modeling of the components will be performed. After that, the transient program code will be made.
*; Yokokawa, Mitsuo
JAERI-M 93-155, 130 Pages, 1993/08
no abstracts in English
*; *; Watanabe, Kenji*; *; *; *; Harada, Hiro; Yokokawa, Mitsuo
JAERI-M 92-105, 209 Pages, 1992/07
no abstracts in English
Yokobayashi, Masao;
Nihon Genshiryoku Gakkai-Shi, 33(7), p.695 - 702, 1991/07
no abstracts in English
Ugajin, Mitsuhiro
JAERI-M 91-069, 54 Pages, 1991/05
no abstracts in English
Kurakami, Junichi; Ouchi, Yuichiro; Yamaki, Takao*
PNC TN8410 91-087, 4 Pages, 1991/03
None
Shinohara, Kunihiko*; Asano, Tomohiro; Narita, Osamu
PNC TN8410 87-17, 40 Pages, 1987/05
A computer code ORION-II has been developed to estimate environmental concentration and dose due to airborne release of radioactive material from multiple sources of nuclear fuel cycle facilities. The modified Gaussian plume model is applied to calculate atmospheric dispersion. The plume depletion models such as gravitational settling, dry deposition, precipitation scavenging and radioactive decay are considered as the causes of the depletion and deposition on the ground and vegetation. Dose to the public is calculated as an individual dose. Inhalation and oral intake are considered as internal pathways. Submersion in the radioactive cloud and external exposure to ground surface contamination are considered as external pathways. ORION-II is the updated version of ORION code (PNCT N841-83-42) and is appricable to the case study of environmental assessment for the nuclear fuel cycle facilities. ORION-II is written in FORTRAN-IV language and can be run on IBM 360, 370, 303X, 308X, 43XX, and FACOM M-series computers.
; Tsunematsu, Toshihide; Takeda, Tatsuoki
JAERI-M 86-159, 32 Pages, 1986/11
no abstracts in English
Takeda, Tatsuoki; Tsunematsu, Toshihide;
JAERI-M 82-097, 64 Pages, 1982/08
no abstracts in English
;
JAERI-M 9719, 71 Pages, 1981/10
no abstracts in English
Nakagawa, Tsuneo
JAERI-M 9499, 54 Pages, 1981/05
no abstracts in English
Nakagawa, Tsuneo
Nihon Genshiryoku Gakkai-Shi, 22(8), p.559 - 571, 1980/00
Times Cited Count：6 Percentile：58.22(Nuclear Science & Technology)no abstracts in English
