Published 1991 | Version v1
Journal article Metadata-only

Analysis of Chinshan-1 power test data with the BNL plant analyzer

Creators

  • 1. National Tsing Hua Univ., Hsinchu (Taiwan)

Description

The Brookhaven National Laboratory (BNL) plant analyzer, also called the high-speed interactive plant analyzer (HIPA), was developed to reduce the time and cost associated with nuclear power plant transient simulation. The BNL analyzer proved to be quite accurate for the analysis of boiling water reactor (BWR) designs and compared well with the results of TRAC-B01, RAMONA-3B, and RELAP-5 calculations. Modification such as feedwater pump control system, SRV logics, input parameter, and form loss coefficient were made to HIPA's BWR/4 models to accommodate the BWR/4 plant-specific design features of Chinshan-1. This paper presents comparisons of HIPA prediction to Chinshan-1 data for two important power tests

Additional details

Publishing Information

Journal Title
Transactions of the American Nuclear Society
Journal Volume
63
Series
Trans. Am. Nucl. Soc.
Journal Page Range
462
ISSN
0003-018X
CODEN
TANSA

Conference

Title
Annual meeting of the American Nuclear Society (ANS).
Dates
2-6 Jun 1991.
Place
Orlando, FL (United States).

INIS

Country of Publication
United States
Country of Input or Organization
United States
INIS RN
23036700
Subject category
S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S42: ENGINEERING;
Resource subtype / Literary indicator
Conference
Descriptors DEI
ACCURACY; B CODES; BNL; BOUNDARY CONDITIONS; CHINSHAN-1 REACTOR; COMPUTER CODES; COMPUTERIZED SIMULATION; CONTROL SYSTEMS; FEEDWATER; FLOW RATE; HEAT TRANSFER; HYDRAULICS; LOSS OF COOLANT; PUMPS; R CODES; REACTOR CORES; REACTOR SAFETY; SCRAM; T CODES; TIME DEPENDENCE; TRANSIENTS
Descriptors DEC
ACCIDENTS; BWR TYPE REACTORS; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; EQUIPMENT; HYDROGEN COMPOUNDS; NATIONAL ORGANIZATIONS; OXYGEN COMPOUNDS; POWER REACTORS; REACTOR ACCIDENTS; REACTOR COMPONENTS; REACTOR SHUTDOWN; REACTORS; SAFETY; SHUTDOWN; SIMULATION; THERMAL REACTORS; US AEC; US DOE; US ERDA; US ORGANIZATIONS; WATER; WATER COOLED REACTORS; WATER MODERATED REACTORS

Optional Information

Secondary number(s)
CONF-910603--.