ML20206A685

Safety Evaluation Supporting Util 861024 Request for Relief from Requirements in Section XI of ASME Boiler & Pressure Vessel Code Re Visual Exam of Accessible Areas of Reactor Pressure Vessel at Approx 3-yr Intervals

ML20206A685
Person / Time
Site:	Fort Calhoun
Issue date:	04/01/1987
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20206A667	List: ML20206A663 ML20206A685
References
TAC-63318, NUDOCS 8704080135
Download: ML20206A685 (4)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT N0. 1 DOCKET NO. 50-285 RELIEF FROM CERTAIN REQUIREMENTS OF SECTION XI 0F THE ASME CODE

1. 0 INTRODUCTION The Technical Specifications for the Fort Calhoun Station, Unit No. 1, require that inservice examination of ASME Code Class 1, 2, and 3 components shall be performed in accordance with Section XI of the ASME Code as required by 10 CFR 50.55a(g)(4) except where specific written relief has been granted by the Commission. Paragraph 10 CFR 50.55a(g)(6)(i) authorizes the Commission to grant relief from those requirements upon making.the necessary findings.

In a letter dated October 24, 1986, Omaha Public Power District (0 PPD), the licensee, identified a specific ASME Code requirement that OPPD determined to be impractical to perform at Fort Calhoun and requested relief from this requirement. The staff has evaluated the licensee's supporting technical 7

justification and finds them to be conditionally acceptable provided that the licensee implements certain alternative requirements.

2.0 EVALUATION OF RELIEF REQUEST The licensee requested relief from a specific inservice inspection (ISI) requirement and provided supporting technical information. The staff reviewed this information as related to the design, geometry, and materials of construction of the components.

Examination Category B-N-1, Item No. 813.10, Interior of the Reactor Vessel.

A. Code Requirements: ASME Section XI, 1980 Edition including Addenda through Winter 1980, requires the following:

A VT-3 visual examination is required of the accessible area:; of the reactor vessel interior at the first refueling outage and subsaquent refueling outages at approximately 3 year intervals. ASME .

Section XI defines the accessible areas as the space above and below the reactor core that is made accessible for examination by removal of components during normal refueling outages.

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Section XI, paragraph IWA-2213, describes the examination method as follows:

(1) The VT-3 visual examination shall be conducted to determine the general mechanical and structural conditions of components and their supports, such as the presence of loose parts, debris, or abnormal corrosion products, wear, erosion, corrosion, and the loss of integrity at bolted or welded connections.

(2) The VT-3 visual examination may require, as applicable to determine structural integrity, the measurement of clearances, detection of physical displacement, structural adequacy of sup-porting elements, connections between load carrying structural members, and tightness of bolting.

(3) For component supports and component interiors, the visual examination may be performed remotely with or without optical aids to verify the structural integrity of the component.

B. Code Relief Request The licensee requests Commission approval to eliminate the VT-3 visual exaainations during the inspection periods (approximately 3 year intervals) in which the core support barrel is not removed. The licensee did not propose an alternative examination.

C. Licensee's Basis For Request The licensee determined that the visual examination defined in Examination Category 8-N-1, Item No. B13.10, with the core support structure installed, is of limited value and results in unnecessary radiation exposure.

Normally, only the uppermost 8 inches of the reactor vessel interior is accessible for visual examination during scheduled plant outages. The components that can be examined are not attached by bolted or welded connections and the probability of detecting loose parts, debris, or abnormal corrosion products, wear, erosion, and corrosion in such a limited area is small. Accessibility is limited by the design of the reactor vessel and is caused by mechanical interference presented by the instrument flange on top of the core support barrel and the reactor vessel internals.

During those plant outages when the core support barrel is removed, the reactor internal surfaces are accessible. Under these condi-tions meaningful information can be obtained by the prescribed visual examinations. Normally, the core support barrel is removed during plant outages corresponding to the end of each 10 year

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interval. The entire reactor vessel interior was visually examined in 1983 during the first 10 year ISI examination. There were no visual indications detected at that time which could be classified as unusual or unacceptable.

It is the opinion of OPPD that the requirement of B13.10 to perform visual examination of accessible reactor vessel internals was not intended for such limited areas as those encountered at Fort Calhoun Station. It is also the opinion of OPPD that the information gained by the prescribed examinations does not warrant the radiation ex-posure to which personnel performing the examinations are subjected.

3 Therefore, OPPD requests. relief from performing the examination required by IWB-2500-1, Category B-N-1, Item B13.10, except for those plant outages when the core support barrel is removed because safety is not enhanced by examining such a small accessible area and because of the radiation exposure to personnel when performing the examination.

D. Staff Evaluation The staff has completed the review of the licensee's letter dated October 24, 1986 based on the provisions of 10 CFR 50.55a(g)(6)(i).

The licensee visually examined the entire reactor vessel interior in 1983 and no visual indications were detected which could be classified as unusual or unacceptable. The staff agrees with the licensee that the visual examination of an 8-inch region of the vessel interior, without bolted or welded connections, will be of limited value. In addition, the licensee probably would detect

-significant degradation, if present, in the region subject to examination during normal refueling operations.

The staff finds that the Code required examinations under Item No. B13.10 can be performed by the licensee. However, in order to eliminate the required visual examinations at the specified frequency, the licensee must propose an alternative inspection program that would provide more meaningful inspection results and, thus, ensure an acceptable level of quality and safety of the reactor vessel interior. The staff determined that performing the VT-3 visual examination of the accessible areas of the reactor vessel interior each time the core support barrel is removed for plant inspection, maintenance, or repair activities would represent an acceptable alternative program. The staff determined that the above alternative inspection program would minimize personnel radiation exposure by performing examinations only when the core

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.o support barrel was scheduled for removal or when known problems were identified in the reactor vessel. The alternative program would provide meaningful results as indicated by the licensee's submittal.

3.0 CONCLUSION

Pursuant to 10 CFR 50.'55a(g)(6)(i) the staff concludes that relief may be granted as requested by the licensee to eliminate the VT-3 visual examination of the reactor vessel interior, except when the core support barrel is removed, provided that the licensee implements an alternative program consisting of performing the Examination Category 8-N-1, Item.No. B13.10 requirement each time the core support barrel is removed.

The staff has-determined that this action is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest, giving due consideration to the burden upon the

. licensee that could result if the code requirements were imposed on the facility.

-Date: April 1,1987 Principal Contributor: '

M. Hum 5

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