Savannah River Site reactor hardware design modification study
A study was undertaken to assess the merits of proposed design modifications to the 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. For the elevated piping design, system recovery was predicted for breaks in the plenum inlet or pump suction piping; response to the pump discharge break location did not show improvement compared to the present system configuration. The rotovalve closure design improved system response to plenum inlet or pump discharge breaks; recovery was not predicted for pump suction breaks. The pump suction valve closure design demonstrated system recovery for all break locations downstream of the valve. A combination of features is recommended to ensure liquid inventory recovery for all break locations. The elevated piping design performance during pump discharge breaks would be improved with addition of a dc pump trip in the affected loop. 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. 12 refs., 10 figs., 2 tabs.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls, ID (United States)
- Sponsoring Organization:
- DOE/NE
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 6913164
- Report Number(s):
- EGG-EAST-8984; ON: DE90010906; TRN: 90-016107
- 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
SAVANNAH RIVER PLANT
HEAT TRANSFER
HYDRAULICS
LEAKS
LOSS OF COOLANT
PIPES
PUMPS
R CODES
REACTOR SAFETY
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