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Nuclear Safety Research Reactor (NSRR) program in JAERI

Conference:

Abstract

An experimental research program, named Nuclear Safety Research Reactor (NSRR) Program, has been progressing in Japan Atomic Energy Research Institute (JAERI) using a modified TRIGA-ACPR. This paper is prepared to describe the outline of the NSRR program. The purpose of the NSRR program is to examine the behaviors of fuel rods under various accidental conditions of power reactors so as to establish realistic safety criteria and to develop analytical models for prediction of fuel failures. We expect to contribute finally to the improvement of reactor design and fuel fabrication techniques based on these experimental results. The NSRR experiments will be performed in the large central experimental tube, which is one of the most excellent features of this reactor, using specially designed capsules or loops which can accommodate up to 49 BWR type test fuels. Many types of test fuels in various conditions will be examined by the NSRR program, such as BWR, PWR and FBR type fuels from the beginning of life to the end of life with and without simulated reactor internal structures. The experiments will be continued for more than 10 years divided into three phases. The first phase of the program will be devoted to the experiments  More>>
Authors:
Ishikawa, M; Hoshi, T; Ohnishi, N; Fujishiro, T; Inabe, T [1] 
  1. Japan Atomic Energy Research Institute (Japan)
Publication Date:
Jul 01, 1974
Product Type:
Conference
Report Number:
INIS-US-09N0328; TOC-5
Resource Relation:
Conference: 3. TRIGA owners' conference, Albuquerque, NM (United States), 25-27 Feb 1974; Other Information: Country of input: International Atomic Energy Agency (IAEA); 7 figs, 3 tabs; Related Information: In: 3. TRIGA owners' conference. Papers and abstracts, 432 pages.
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; BWR TYPE REACTORS; DOLLARS; FUEL RODS; HEAT; HEAT FLUX; JAERI; LOSS OF COOLANT; NUCLEAR FUELS; POWER-COOLING-MISMATCH ACCIDENTS; PWR TYPE REACTORS; RADIATION PROTECTION; REACTIVITY; REACTIVITY INSERTIONS; REACTOR PERIOD; RESEARCH PROGRAMS; RESEARCH REACTORS; URANIUM DIOXIDE
OSTI ID:
21217738
Research Organizations:
General Atomic Co., San Diego, CA (United States)
Country of Origin:
United States
Language:
English
Other Identifying Numbers:
TRN: US09N0345086522
Availability:
Available from INIS in electronic form
Submitting Site:
INIS
Size:
page(s) 3.1-3.14
Announcement Date:
Oct 17, 2009

Conference:

Citation Formats

Ishikawa, M, Hoshi, T, Ohnishi, N, Fujishiro, T, and Inabe, T. Nuclear Safety Research Reactor (NSRR) program in JAERI. United States: N. p., 1974. Web.
Ishikawa, M, Hoshi, T, Ohnishi, N, Fujishiro, T, & Inabe, T. Nuclear Safety Research Reactor (NSRR) program in JAERI. United States.
Ishikawa, M, Hoshi, T, Ohnishi, N, Fujishiro, T, and Inabe, T. 1974. "Nuclear Safety Research Reactor (NSRR) program in JAERI." United States.
@misc{etde_21217738,
title = {Nuclear Safety Research Reactor (NSRR) program in JAERI}
author = {Ishikawa, M, Hoshi, T, Ohnishi, N, Fujishiro, T, and Inabe, T}
abstractNote = {An experimental research program, named Nuclear Safety Research Reactor (NSRR) Program, has been progressing in Japan Atomic Energy Research Institute (JAERI) using a modified TRIGA-ACPR. This paper is prepared to describe the outline of the NSRR program. The purpose of the NSRR program is to examine the behaviors of fuel rods under various accidental conditions of power reactors so as to establish realistic safety criteria and to develop analytical models for prediction of fuel failures. We expect to contribute finally to the improvement of reactor design and fuel fabrication techniques based on these experimental results. The NSRR experiments will be performed in the large central experimental tube, which is one of the most excellent features of this reactor, using specially designed capsules or loops which can accommodate up to 49 BWR type test fuels. Many types of test fuels in various conditions will be examined by the NSRR program, such as BWR, PWR and FBR type fuels from the beginning of life to the end of life with and without simulated reactor internal structures. The experiments will be continued for more than 10 years divided into three phases. The first phase of the program will be devoted to the experiments pertaining to reactivity initiated accidents (RIA). In these experiments we will make use of the excellent pulsing capability of ACPR, which is expected to generate 100 MW-sec prompt energy release with 1.3 msec of minimum reactor period by 4.7 dollar reactivity insertion and to yield more than 280 cal/g-UO{sub 2} heat deposit even in an approximately 10% enriched BWR type test fuel. (280 cal/g-UO{sub 2} is believed enough heat deposit to cause fuel failure.) In general, heat flow behaviors from fuel meat to clad and from clad to coolant are very complex phenomena, but they are the key point in analyzing transient response of fuels. In the sudden heat transient conditions brought by pulsing, however, it will be possible to examine each phenomenon separately according to their characteristics of time delay. This is the main reason why RIA type experiments are selected at the initial stage of the NSRR program. The second phase of the program will be devoted to the experiments pertaining to power cooling mismatch (PCM) accidents and possibly loss of coolant accidents (LOCA). In these experiments, we will make use of the steady state power level which will be increased up to 14 MW. The third phase will be the experiments pertaining to burned-up fuels. (author)}
place = {United States}
year = {1974}
month = {Jul}
}