skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Savannah River Site reactor hardware design modification study

Abstract

A study was undertaken to assess the merits of proposed design modifications to the Savannah River Site (SRS) reactors. The evaluation was based on the responses calculated by the RELAP5 systems code to double-ended guillotine break loss-of-coolant-accidents (DEGB LOCAs). The three concepts evaluated were (a) elevated plenum inlet piping with a guard vessel and clamshell enclosures, (b) closure of both rotovalves in the affected loop, and (c) closure of the pump suction valve in the affected loop. Each concept included a fast reactor shutdown (to 65% power in 100 ms) and a 2-s ac pump trip. System recovery potential was evaluated for break locations at the pump suction, the pump discharge, and the plenum inlet. The code version used was RELAP5/MOD2.5 version 3d3, a preliminary version of RELAP5/MOD3. The model was a three-dimensional representation of the K-Reactor water plenum and moderator tank. It included explicit representations of all six loops, which were based on the configuration of L-Reactor. A combination of features is recommended to ensure liquid inventory recovery for all break locations. Valve closure design performance for a break location in the short section of piping between the reactor concrete shield and the pump suction valve would benefit frommore » the clamshell enclosing that section of piping. 7 refs., 10 figs., 2 tabs.« less

Authors:
Publication Date:
Research Org.:
EG and G Idaho, Inc., Idaho Falls, ID (USA)
Sponsoring Org.:
DOE/NE
OSTI Identifier:
6346024
Report Number(s):
EGG-M-90342; CONF-9009219-3
ON: DE91001834; TRN: 90-035996
DOE Contract Number:  
AC07-76ID01570
Resource Type:
Conference
Resource Relation:
Conference: 1990 joint RELAP5 and TRAC-BWR international user seminar, Chicago, IL (USA), 17-21 Sep 1990
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; PRODUCTION REACTORS; REACTOR COOLING SYSTEMS; MODIFICATIONS; DESIGN; HEAT TRANSFER; HYDRAULICS; LOSS OF COOLANT; PIPES; R CODES; REACTOR SAFETY; SAVANNAH RIVER PLANT; VALVES; ACCIDENTS; COMPUTER CODES; CONTROL EQUIPMENT; COOLING SYSTEMS; ENERGY SYSTEMS; ENERGY TRANSFER; EQUIPMENT; FLOW REGULATORS; FLUID MECHANICS; MECHANICS; NATIONAL ORGANIZATIONS; REACTOR ACCIDENTS; REACTOR COMPONENTS; REACTORS; SAFETY; US AEC; US DOE; US ERDA; US ORGANIZATIONS; 220900* - Nuclear Reactor Technology- Reactor Safety; 220700 - Nuclear Reactor Technology- Plutonium & Isotope Production Reactors

Citation Formats

Fisher, J E. Savannah River Site reactor hardware design modification study. United States: N. p., 1990. Web.
Fisher, J E. Savannah River Site reactor hardware design modification study. United States.
Fisher, J E. 1990. "Savannah River Site reactor hardware design modification study". United States. https://www.osti.gov/servlets/purl/6346024.
@article{osti_6346024,
title = {Savannah River Site reactor hardware design modification study},
author = {Fisher, J E},
abstractNote = {A study was undertaken to assess the merits of proposed design modifications to the Savannah River Site (SRS) reactors. The evaluation was based on the responses calculated by the RELAP5 systems code to double-ended guillotine break loss-of-coolant-accidents (DEGB LOCAs). The three concepts evaluated were (a) elevated plenum inlet piping with a guard vessel and clamshell enclosures, (b) closure of both rotovalves in the affected loop, and (c) closure of the pump suction valve in the affected loop. Each concept included a fast reactor shutdown (to 65% power in 100 ms) and a 2-s ac pump trip. System recovery potential was evaluated for break locations at the pump suction, the pump discharge, and the plenum inlet. The code version used was RELAP5/MOD2.5 version 3d3, a preliminary version of RELAP5/MOD3. The model was a three-dimensional representation of the K-Reactor water plenum and moderator tank. It included explicit representations of all six loops, which were based on the configuration of L-Reactor. A combination of features is recommended to ensure liquid inventory recovery for all break locations. Valve closure design performance for a break location in the short section of piping between the reactor concrete shield and the pump suction valve would benefit from the clamshell enclosing that section of piping. 7 refs., 10 figs., 2 tabs.},
doi = {},
url = {https://www.osti.gov/biblio/6346024}, journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Jan 01 00:00:00 EST 1990},
month = {Mon Jan 01 00:00:00 EST 1990}
}

Conference:
Other availability
Please see Document Availability for additional information on obtaining the full-text document. Library patrons may search WorldCat to identify libraries that hold this conference proceeding.

Save / Share: