Savannah River Site reactor hardware design modification study
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.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls, ID (USA)
- Sponsoring Organization:
- DOE/NE
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 6346024
- Report Number(s):
- EGG-M-90342; CONF-9009219-3; ON: DE91001834; TRN: 90-035996
- 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
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Related Subjects
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