STP909

    Investigations and Measures to Guarantee the Safety of a Pressurized Water Reactor's Pressure Vessel Against Brittle Fracture

    Published: Jan 1986


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    Abstract

    This paper presents the investigations and procedures being taken to guarantee the integrity of reactor pressure vessels (RPV) as a result of radiation embrittlement of the weld seam in the core region (22 NiMoCr 3 7 corresponding to ASTM A508, Class 2).

    First, changes in the core map substantially reduced radiation exposure rates for the vessel wall. Second, experimental radiation programs on specimens and measured fracture toughness, Kk, of the welding material, as a function of neutron flux, are discussed together with the theoretical and experimental determination of neutron flux. As the high-pressure injection system is modified to hot leg injection, a depressurized thermal shock produces the critical pressure and temperature transients, which covers all other thermo-hydraulic plant transients with respect to brittle fracture loads.

    Additionally, analytical results for the initiation and arrest of circumferential cracks, warm prestress effects for the end-of-life (EOL) state of the RPV, and limited crack size investigations are discussed. Finally, plant inspection plans with respect to vessel wall radiation exposure and nondestructive testing are illustrated.

    Keywords:

    radiation embrittlement, radiation exposure program, thermoshock, brittle fracture safety, crack initiation, crack arrest analysis, radiation effects, pressure vessel steels, welded joints


    Author Information:

    Kaun, UE
    Division general manager and supervisor, TUV Norddeutschland, Hamburg,

    Koehring, FKA
    Division general manager and supervisor, TUV Norddeutschland, Hamburg,


    Paper ID: STP23033S

    Committee/Subcommittee: E10.07

    DOI: 10.1520/STP23033S


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