Community Research and Development Information Service - CORDIS

Abstract

The operability of Pressurized Water Reactor (PWR) plants is largely influenced by hypothetical Pressurized Thermal Shock (PTS) loads on the pressure vessel walls. Any assessment of PTS severity requires knowledge of the thermal-hydraulic flow mixing phenomena within the reactor pressure vessel downcomer in order to quantify the metallurgical and fracture mechanics aspects of crack growth and propagation. This paper focuses on certain thermal-hydraulic aspects of the issue which have a direct relevance for the development of analytical models and the assessment of system codes used in the safety evaluation of large PWRs. Specific reference is made to the interaction of high pressure injection system emergency core coolant flow with primary loop flow and the influence thereof on vessel thermal hydraulics in small break loss of coolant experiments performed in the LOBI-MOD2 test facility.

Additional information

Authors: ADDABBO C, JRC ISPRA ESTAB.(ITALY);PIPLIES L, JRC ISPRA ESTAB.(ITALY);WORTH B JRC ISPRA ESTAB.(ITALY), JRC ISPRA ESTAB.(ITALY)
Bibliographic Reference: ANS/ENS TOPICAL MEETING ON "OPERABILITY OF NUCLEAR POWER SYSTEMS IN NORMAL AND ADVERSE ENVIRONMENT", ALBUQUERQUE, NEW MEXICO (USA), SEPT. 29-OCT. 2, 1986 WRITE TO CEC LUXEMBOURG, DG XIII/A2, POB 1907 MENTIONING PAPER E 32908 ORA
Availability: Can be ordered online
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