Published April 1996 | Version v1
Report Open

COAST code conversion from Cyber to HP

Creators

  • 1. Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

Description

The transient thermal hydraulic behavior of reactor coolant system in a nuclear power plant following loss of coolant flow is analyzed by use of COAST digital computer code. COAST calculates individual loop flow rates and steam generator pressure drops is a function of time following coast-down of any number of reactor coolant pumps. This report firstly describes detailed work carried out for installation of COAST on HP 9000/700 series and code validation results after installation. Secondly, a series of work is also describes in relation to installation of COAST on Apollo DN10000 series as well as relevant code validation results. Attached is a report on software verification and validation results. 7 refs. (Author) .new

Availability note (English)

MF available from INIS under the Report Number.

Files

27063045.pdf

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Additional details

Publishing Information

Imprint Pagination
34 p.
Report number
KAERI/TR--660/96

INIS

Country of Publication
Korea, Republic of
Country of Input or Organization
Korea, Republic of
INIS RN
27063045
Subject category
S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S99: GENERAL AND MISCELLANEOUS; S22: GENERAL STUDIES OF NUCLEAR REACTORS;
Descriptors DEI
COMPUTER CODES; HYDRAULICS; LOSS OF COOLANT; NUCLEAR POWER PLANTS; PUMPS; STEAM GENERATORS; THERMODYNAMICS
Descriptors DEC
ACCIDENTS; BOILERS; EQUIPMENT; NUCLEAR FACILITIES; POWER PLANTS; REACTOR ACCIDENTS; THERMAL POWER PLANTS; VAPOR GENERATORS