https://re.public.polimi.it/bitstream/11311/1020691/1/rev_A%20Dynamic%20safety%20margin%20characterization%20approach%20in%20support%20of%20IDPSA.pdf

A dynamic probabilistic safety margin characterization approach in support of Integrated Deterministic and Probabilistic Safety Analysis

Since the Three Mile Island accident, an important focus of pressurized water reactor (PWR) transient analyses has been a small-break loss-of-coolant accident (SBLOCA). In 2002, the discovery of th...

/pdf/accident-management-actions-in-an-upper-head-small-break-107hjlbjx5.pdf

Accident Management Actions in an Upper-Head Small-Break Loss-of-Coolant Accident with High-Pressure Safety Injection Failed

Effects of RCP trip when recovering HPSI during LOCA in a Westinghouse PWR

Instrumentation needs for the detection of inadequate core cooling are discussed, and a review of reactor coolant inventory tracking systems (RCITS) is presented. RCITS installation experience is considered, and attention is given to operating experience, accuracy, and procedures. It is noted that the addition of inadequate core cooling accident monitoring instrumentation to all PWRs (pressurized water reactors) is nearing completion after more than eight years. The response of and acceptance by the utilities have been largely positive, with only a few holdouts. It is concluded that the systems developed, while not providing the precision that might be desired, are sufficient to provide vital information for accident avoidance and recovery. In addition, unexpected benefits for normal operation have resulted, and operator acceptance and confidence have been good. >

/pdf/post-implementation-review-of-inadequate-core-cooling-3lov5ab9uv.pdf

Post-implementation review of inadequate core cooling instrumentation

Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary sides in case of core temperature excursion. Test 6-1 is the first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reasons of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection.Copyright © 2008 by ASME

Performance of Core Exit Thermocouple for PWR Accident Management Action in Vessel Top Break LOCA Simulation Experiment at OECD/NEA ROSA Project

Instrumentation needs for detection of inadequate core cooling. Studies of the Three Mile Island accident identified the need for additional instrumentation to detect inadequate core cooling (ICC) in nuclear power plants. Industry studies by plant owners and reactor vendors supported the conclusion that improvements were needed to help operators diagnose the approach to or existence of ICC as well as to provide more complete information for operator control of safety injection flow to minimize the consequences of such an accident. In 1980, the US Nuclear Regulatory Commission (NRC) required further studies by the industry and described ICC instrumentation design requirements that included human factors and environmental considerations. On December 10, 1982, NRC issued to Babcock and Wilcox (B and W) licensees orders for Modification of License and transmitted to pressurized water reactor licensees Generic Letter 82-28 to inform them of the revised NRC requirements. The instrumentation requirements include upgraded subcooling margin monitors (SMM), upgraded core exit thermocouples (CET), and installation of a reactor coolant inventory tracking system. NRC Regulatory Guide 1.97, which covers accident monitoring instrumentation, was revised (Rev. 3) to be consistent with the requirements of item II.F.2 of NUREG-0737.

L.E. Phillips

Papers

Post-implementation review of inadequate core cooling instrumentation

Instrument accuracy in reactor vessel inventory tracking systems