Analysis of the source range monitor during the first four hours of the Three Mile Island Unit 2 accident

Journal Article · Wed Feb 01 00:00:00 EST 1989 · Nuclear Technology; (USA)

OSTI ID:5678361

Wu, H Y; Bandini, B R ^[1]; Hsiao, M Y; Baratta, A J; Bandini, B R ^[2]; Tolman, E L ^[3]

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1. Institute of Nuclear Energy Research, Lungtan (TW)
2. Pennsylvania State Univ., University Park, PA (USA). Dept. of Nuclear Engineering
3. Idaho National Engineering Lab., Idaho Falls, ID (USA)

The source range monitor (SRM) data recorded during the first 4 h of the Three Mile Island Unit 2 (TMI-2) accident following reactor shutdown were analyzed. An effort to simulate the actual SRM response was made by performing a series of neutron transport calculations. Primary emphasis was placed on simulating the changes in SRM response to various system events during the accident so as to obtain useful information about core conditions at the various stages. Based on the known end-state reactor conditions, the major system events and the actual SRM readings, self-consistent estimates were made of core liquid level, void fraction in the coolant, and locations of core materials. This analysis expands the possible interpretation of the SRM data relative to core damage progression. The results appear to be consistent with other studies of the TMI-2 Accident Evaluation Program, and provide information useful for the development and determination of the TMI-2 accident scenario.

• OSTI ID: 5678361
• Journal Information: Nuclear Technology; (USA), Vol. 84:2; ISSN 0029-5450
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
NEUTRON TRANSPORT
CALCULATION METHODS
REACTOR CORES
DAMAGE
THREE MILE ISLAND-2 REACTOR
REACTOR ACCIDENTS
REACTOR SAFETY EXPERIMENTS
ACCIDENTS
ENRICHED URANIUM REACTORS
NEUTRAL-PARTICLE TRANSPORT
POWER REACTORS
PWR TYPE REACTORS
RADIATION TRANSPORT
REACTOR COMPONENTS
REACTORS
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
220100 - Nuclear Reactor Technology- Theory & Calculation