The Japan Atomic Energy Research Institute (JAERI) and the United States Nuclear Regulatory Commission (USNRC) are jointly conducting confirmatory, integral testing on the Westinghouse AP600 reactor transient responses by using the ROSA-V Large Scale Test Facility of JAERI. This facility, built originally to simulate conventional 4-loop pressurized water reactors (PWRS) , has been modified by adding components specific to the AP600 design. The modified LSTF now provides a full-pressure, full-height, 1/30.5 volumetrically-scaled simulation of AP600. Five loss-of-coolant accident (LOCA) experiments were performed by August 1994, simulating transients initiated by cold leg breaks, a Pressure Balance Line (PBL) break, and inadvertently open Automatic Depressurization System (ADS) valves. These experiments indicated adequate core cooling and decay heat removal performance of the AP600 passive safety components.
i A rising level of' scrutiny is being directed toward the Savannah River Site (SRS) production reactors. Improved calculational capabilities are being developed to provide a best estimate analytical process to determine the safe operating margins of the reactors. The Code Scaling, Applicability, and Uncertainty (CSAU) methodology, developed by the U. S. Nuclear Regulatory Commission to support best estimate simulations, is being applied to the best estimate limits analysis for the SRS production reactc-s. One of the foundational parts of the method is the identification and ranking of ali the processes that occur during the specific limiting scenario. The phenomena ranking is done according to their importance to safety criteria during the transient and is used to focus the tmcerminty analysis on a sufficient, yet cost effective scope of work. This report documents the thermal-hydraulic phenomena that occur during a limiting break in an SRS production reactor and their importance to the uncertainty in simulations of the reactor behavior.ii II SUMMARY .0The _ scrutiny directed toward the operation of Department of Energy production reactors in recent years has led to the development and incorporationof best-estimate computer codes in the safety analysis process. The use of best-estimate techniques requires that the analysis be accompanied by a quantification of the uncertainty in the calculated restdts. The operating power limit for the SRS production reactors is determined through the use of computer codes. The CSAU methodology, developed by the U. S. Nuclear Regulatory Commission to support best estimate analyses for light water reactors,is being applied to the power limiting transientfor the SRS production reactors.The first segment in _e CSAU methodology is the identification and ranking of phenomena that are imtxn'tant to the limiting scenario. Since it is not cost effective to assess ali models in the code the CSAU method provides justification for investigating only the important phenomena. The selection is made according to a ranking of the importance the phenomena have with respect to safe reactor operations. The purpose of this report is to identify the thermal-hydraulic phenomena associated with the limiting break in an SRS production reactor and their importance to the safety criteria used to establish acceptable safety margins.
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