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DOE OSTI2012

Spent fuel pool analysis using TRACE code

The storage requirements of Spent Fuel Pools have been analyzed with the purpose to increase their rack capacities. In the past, the thermal limits have been mainly evaluated with conservative codes developed for this purpose, although some works can be found in which a best estimate code is used. The use of best estimate codes is interesting as they provide more realistic calculations and they have the capability of analyzing a wide range of transients that could affect the Spent Fuel Pool. Two of the most representative thermal-hydraulic codes are RELAP-5 and TRAC. Nowadays, TRACE code is being developed to make use of the more favorable characteristics of RELAP-5 and TRAC codes. Among the components coded in TRACE that can be used to construct the model, it is interesting to use the VESSEL component, which has the capacity of reproducing three dimensional phenomena. In this work, a thermal-hydraulic model of the Maine Yankee spent fuel pool using the TRACE code is developed. Such model has been used to perform a licensing calculation and the results obtained have been compared with experimental measurements made at the pool, showing a good agreement between the calculations predicted by TRACE and the experimental data. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

Effect of steam generator configuration in a loss of the RHR during mid-loop operation at PKL facility

The loss of the residual heat removal system in mid-loop conditions may occur with a non-negligible contribution to the plant risk, so the analysis of the accidental sequences and the actions to mitigate the accident are of great interest in shutdown conditions. In order to plan the appropriate measures to mitigate the accident is necessary to understand the thermal-hydraulic processes following the loss of the residual heat removal system during shutdown. Thus, transients of this kind have been simulated using best-estimate codes in different integral test facilities and compared with experimental data obtained in different facilities. In PKL (Primaerkreislauf-Versuchsanlage, primary coolant loop test facility) test facility different series of experiments have been undertaken to analyze the plant response in shutdown. In this context, the E3 and F2 series consist of analyzing the loss of the residual heat removal system with a reduced inventory in the primary system. In particular, the experiments were developed to investigate the influence of the steam generators secondary side configuration on the plant response, what involves the consideration of different number of steam generators filled with water and ready for activation, on the heat transfer mechanisms inside the steam generators U-tubes. This work presents the results of such experiments calculated using, RELAP5/Mod 3.3. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗