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.

This report documents work performed at the Idaho National Engineering Laboratory (INEL) in support of Westinghouse Hanford Company safety analyses for the N Reactor. The portion of the work reported here includes comparisons of RELAP5/MOD2-calculated data with measured plant data for^ (1) a plant trip reactor transient from full power operation, and (2) a hot dump test performed prior to plant startup. These qualitative comparisons are valuable because they provide an indication of the capabilities of the RELAP5/M0D2 code to simulate operational and blowdown transients in the N Reactor.

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