Resolution of Generic Safety Issues: Item B-47: Inservice Inspection of Supports - Classes 1, 2, 3, and MC Components ( 2) ( NUREG-0933, Main Report with Supplements 1–35 )
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DESCRIPTION
According to NUREG-0471,3 the safety concern in this issue relates to results from inspections of various structural components in the torus support systems of operating BWRs. These inspections have revealed several inconsistencies between the design drawings and the as-built hardware, including missing support struts, out-of-tolerance weld dimensions, unwelded regions, and unsupported columns. In addition, a limited number of separate inspections have been performed on PWR steam generator supports and the results of these inspections revealed several cracked support bolts. It is also noted here that several similar problems were identified in late 1981 for the Diablo Canyon nuclear plant. The NRC identified apparent errors in the seismic design resulting from a discrepancy in documents used for the seismic analysis of the plant which led to improper mounting of pipe hanger supports.
NUREG-04713 states that additional investigation of BWR and PWR component support systems should be undertaken in order to determine if similar deficiencies and "off-design" conditions exist generally in operating plants. It is also stated that this investigation should determine the extent of system support deficiencies and whether the deficiencies are service-induced or are the result of faulty construction. The determination of the deficiencies is necessary in order to define the possible safety significance and to provide guidance for further appropriate staff action regarding ISI of supports.
A related study is being conducted under Item A-12, "Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports." In addition, the ASME Boiler and Pressure Vessel Code, Section XI,14 Rules for Inservice Inspection of Nuclear Power Plant Components--Division 1, also applies to this issue. A technical assistance program is presently being funded at INEL by RES to develop data on the possible impact of the proposed exemption of supports from examination included in Subcommittee Section XI Code agenda Item ISI-79-072.
CONCLUSION
Two principal areas of safety concern are identified in this issue: (1) discrepancies between the design drawings and the as-built hardware, and (2) degradation of component supports. Regarding the discrepancies between the support design drawings and the as-built support hardware, these are problems that related directly to the quality assurance program of the licensee and its contractors existing during construction and are not part of the ISI program per se. Therefore, these deficiencies are failures to implement the QA program. No changes in QA criteria of requirements are indicated.
With regard to the degradation of supports, the ASME Code, Section XI14 (1980 edition), addresses the matter of ISI of component supports for Classes 1, 2, 3, and MC components (Subsection IWF) and contains the inservice requirements which appear to fully address the concerns in this issue. Moreover, the current effort under Item A-12 will result in a NUREG document in which guidance and requirements for the selection of materials and the construction of reactor coolant pump and steam generator support structures will be addressed. In addition, preservice and inservice inspection requirements of these support structures for operating plants will also be addressed in this NUREG.
In view of the existing ISI requirements for supports of Classes 1, 2, 3, and MC components and QA program requirements, it appears that the concerns in this issue are already being addressed and that no additional safety benefit could be expected from this issue as stated. Therefore, this item should be DROPPED from further consideration.
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