Assessment of typical BWR (boiling water reactor) vessel configurations and examination coverage

Technical Report · Mon Oct 01 00:00:00 EDT 1990

OSTI ID:6530860

Walker, S M ^[1]; Feige, E J; Ingamells, J R ^[2]; Calhoun, G L; Davis, J; Kapoor, A ^[3]

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1. EPRI Nondestructive Evaluation Center, Charlotte, NC (USA)
2. Southwest Research Inst., San Antonio, TX (USA)
3. Westinghouse Electric Corp., Pittsburgh, PA (USA)

Even though boiling water reactors (BWRs) are not susceptible to the kind of incident known as pressurized thermal shock that must be considered in the design and operation of pressurized water reactors, BWR reactor pressure vessels (RPVs) have experienced higher than expected embrittlement caused by fast neutron irradiation. This has required the vessel to be at a higher temperature than originally projected before the plant can be taken to power operation. In addition, many BWR plants have received exemption from the 10-year volumetric nondestructive evaluation (NDE) of the vessel as required by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B PV) Code Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components,'' because NDE access is severely restricted. Since many RPV welds have not been examined after being placed in service and the potential for service-induced flaws exists, regulatory authorities are looking closely at examination relief requests. BWR reactor vessel examination coverage is typically limited by plant design. Most BWR plants were designed when inservice examination codes were in the early stages of development, and very little consideration was give to designing for NDE access. Consequently, there is restricted access for many areas of the RPV. Since an increase in examination requirements has been placed in ASME B PV Code Section XI in these areas, efforts have begun on a thorough analysis of the vessel weld volumes examined during inservice examination and an evaluation of possibility expanding the RPV examination coverage. Because of these concerns, an investigation of the accessibility of the reactor vessel for NDE was performed to define the present status and to determine the improvements in coverage that can be accomplished in the near future. 7 refs., 9 figs., 4 tabs.

• Research Organization: Electric Power Research Inst., Palo Alto, CA (USA); Jones (J.A.) Applied Research Co., Charlotte, NC (USA); EPRI Nondestructive Evaluation Center, Charlotte, NC (USA); Southwest Research Inst., San Antonio, TX (USA); Westinghouse Electric Corp., Pittsburgh, PA (USA)
• Sponsoring Organization: EPRI
• OSTI ID: 6530860
• Report Number(s): EPRI-NP-7026-D; TRN: 90-035221
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
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PRESSURE VESSELS
IN-SERVICE INSPECTION
NONDESTRUCTIVE ANALYSIS
BRUNSWICK-2 REACTOR
DESIGN
DRESDEN-2 REACTOR
MILLSTONE-1 REACTOR
RECOMMENDATIONS
ULTRASONIC TESTING
VERMONT YANKEE REACTOR
WELDED JOINTS
ACOUSTIC TESTING
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ENRICHED URANIUM REACTORS
INSPECTION
JOINTS
MATERIALS TESTING
NONDESTRUCTIVE TESTING
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WATER COOLED REACTORS
WATER MODERATED REACTORS
210100* - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled