Development of a LOCA analysis code for the supercritical-pressure, light-water-cooled reactors

Book · Thu Aug 01 04:00:00 EDT 1996

OSTI ID:271888

Lee, J H; Koshizuka, S; Oka, Y ^[1]

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1. Univ. of Tokyo, Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.

The supercritical-pressure, light-water-cooled reactors can achieve the cost reduction by system simplification and higher thermal efficiency, and has the flexibility for fuel cycle owing to the technical feasibility of various neutron spectrum reactors. Since the LOCA behavior of the reactor at supercritical pressure condition can not be analyzed with the developed codes for the current light water reactors (LWRs), the LOCA analysis code for the supercritical-pressure, light-water-cooled reactor is developed in this study. The presented code is composed of two parts; the blowdown and the reflood analysis module. The blowdown analysis code is designed based on the homogeneous equilibrium model and the reflood analysis code is modeled by the thermal equilibrium relative velocity model. The code is validated with the REFLA-TRAC code, which is developed in the Japan Atomic Energy Research Institute based on TRAC-PF1. The large break LOCA of the thermal neutron spectrum reactor (SCLWR) is analyzed. The result shows that the maximum peak cladding temperature (PCT) is nearly 980 C at about 60 sec after the break and the PCT position is started to be quenched at 170 sec. This means that the maximum PCT is sufficiently lower than the safety limit of fuel cladding temperature, 1,260 C. In conclusion, the developed code can conform the safety of SCLWR under the LOCA, and is expected to be usefully applied in the LOCA analysis of the other types of supercritical-pressure, light-water-cooled reactors.

• OSTI ID: 271888
• Report Number(s): CONF-960306--; ISBN 0-7918-1226-X
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
99 GENERAL AND MISCELLANEOUS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
S CODES
SAFETY ANALYSIS
THEORETICAL DATA
VALIDATION
WATER COOLED REACTORS