ML20248B962

Safety Evaluation Supporting Amend 109 to License DPR-54

ML20248B962
Person / Time
Site:	Rancho Seco
Issue date:	06/05/1989
From:	Office of Nuclear Reactor Regulation
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ML20248B950	List: ML20248B945 ML20248B953 ML20248B962
References
NUDOCS 8906090220
Download: ML20248B962 (4)

Text

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\,...../ l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0.109TO FACILITY OPERATING LICENSE DPR-54 l RANCHO SEC0 NUCLEAR GENERATING STATION, UNIT 1 DOCKET NO. 50-312

1.0 INTRODUCTION

On September 24, 1986 the Sacramento Municipal Utility District (the Licensee) submitted a proposed amendment to its Facility Operating License DPR-54 for Rancho Seco Nuclear Generating Station Unit No. 1.

The proposed amendment would revise various parts of Specification 4.17 (Steam Generators) to add sleeving as a method of mitigating tube leaks.

This change would allow using a Babcock and Wilcox mechanical sleeving methodology for Rancho Seco's steam generator tube repairs if needed.

By letter dated April 7, 1989 the NRC staff requested that the Sacramento Municipal Utility District make specific changes to their proposed Amendment No. 134 that would be consistent with current NRC guidance. On April 27, 1989 the Sacramento Municipal Utility District submitted information to the NRC which updated their previous request.

2.0 BACKGROUND

Rancho Seco Nuclear Generating Station, a pressurized water reactor, has two steam generators manufactured by Babcock and Wilcox. Each of the two once-through steam generators has over 15,000 Inconel 600 straight tubes which terminate in a tube sheet at both ends of the steam generator, These tubes have an outside diameter of approximately 0.625 inch e d a wall thickness of 0.035 inch.

Rancho Seco 1 started commercial operation in April 1975 and uses all volatile water treatment for its secondary water chemistry. Starting in May 1981 when the plant was shut down due to a large,10 gpm leak, l Rancho Seco has been experiencing tube leaks. These leaks have been l

occurring at the 15th support p' late elevation in tubes adjacent to the open inspection lane.

Tube leakage has been attributed to circumferential through-wall fatigue cracks. The crack initiation is believed to be caused by concentrated chemical contaminants which are carried by moisture in the steam flowing up through the open tube lane. Once the crack is chemically initiated on the outside surface of the tube, it propagates through the wall and then continues circumferentially in both directions around the tube by a high-frequency low-stress fatigue mechanism.

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3.0 DISCUSSION By letter dated September 24, 1986 the Sacramento Municipal Utility District submitted a proposal to amend its operating license for Rancho Seco Nuclear Generating Station Unit No. 1.

The proposed amendment would revise various parts of Specification 4.17 (Steam Generators) to add sleeving as a method of mitigating once-through steam generator leaks. _ Basically, the wording " plug defective tubes" is changed throughout Specification 4.17 and its Bases to " plug or repair defective tubes." In addition, Specification 4.17.2.b.2 revises the first sample inspection to exclude tubes with previous detectable wall penetrations if they have been plugged or sleeve repaired in the affected area. These changes are proposed to permit the use of steam generator tube sleeving as an alternate to tube described in Babcock and Wilcox (B&W) plugging.

report, The sleeving BAW-1823P, Revision methodology 1,

"Once-Through Steam Generator Mechanical Sleeve Qualification" dated November 1985, will be used.

The sleeves are fabricated of Inconel 600 which has been given a thermal treatment to improve its corrosion resistance. The sleeves would be installed in the top portion of the steam generator tube (intersecting the 15th support plate and the upper tube sheet).

The B&W report addressed pertinent aspects of the sleeving which affect nuclear safety such as strength and leakage of the mechanical sleeves, corrosion resistance, effect on plant performance, vibration, method of installation, examination for defects and verifications of proper installa-tion, radiological exposure and analysis for transient and accident conditions. The report concludes that the mechanical tube sleeves are qualified for use in degraded tubes and are strong enough and sufficiently leak-free to significantly extend the life ot degraded tubes. It also recommends that up to 10,000 of these mechanical sleeves can be installed in the steam generators to correct or prevent tube degradation. The April 27, 1989 submittal by the licensee updated the previously requested technical specification changes in order to be consistent with NRC guidance.

In this submittal, clarifications were added regarding: (a) unacceptability of analysis allowing imperfections greater than 40% of the nominal tube

, or sleeve wall thickness, (b) applicability of inspection requirement for l all steam generator tubes, including those previously repaired by sleeving,

[ (c) limitation of sleeving as the method of repair to not more than 5,000

( tubes per steam generator, and (d) reference document for sleeving method-ology.

Four units in two other plants have been given approval by the NRC to use the B&W sleeve process to repair steam generator tubes - Oconee Units 1, 2, 3 and Arkansas Nuclear One, Unit 1. Service experience with the .

sleeving process has been favorable. The staff has reviewed the licensee's 1 proposed technical specification changes and the applicability of the referenced sleeving process to the Rancho Seco plant and finds that they l are acceptable.

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. If improved sleeving techniques are developed in the future, the licensee may use the improved methodology provided that the methodology is reviewed and approved by the NRC staff for use at the Rancho Seco plant. The requirement for staff approval of sleeving methodologies was added to the licensee's submittal by the NRC staff to make the specification more flexible and to take advantage of future technological advances. The NRC initiated change consists of addition of the last sentence to Specification 4.17.4.4 and deletion of reference to the BAW-1823P, Rev. I sleeving methodology in the bases section of the specification. The NRC initiated change was communicated to the licensee and the licensee concurred that the change was desirable.

In addtion, by letter dated May 26, 1989, the licensee provided a commit-ment that it would continue to employ technologically advanced methods of inservice inspection, in accordance with industry accepted methods of eddy current technology in order to increase confidence in the ability to accurately detect steam generator tube and tube / sleeve degradation.

In summary, the staff finds that the steam generator tube repair method by sleeving described in Babcock and Wilcox Report BAW-1823P Revision 1, "Once-Through Steam Generator Mechanical Sleeve Qualification," November 1985, together with the licensee's commitment to utilize technologically advanced methods of inspection in order to increase confidence in its ability to detect tube and tube / sleeve degradation is acceptable as an alternative to the repair by pluggir.g of defecti/e tubes.

4.0 CONTACT WITH STATE OFFICIAL The NRC staff has has advised the Chief of the Radiological Health Branch, State Department of Health Services, State of California, of the proposed determination of no significant hazards consideration. No comments were received.

5.0 ENVIRONMENTAL CONSIDERATION

This amendment involves changes in the installation or use of a facility component located within the restricted area as defined in 10 CFR 20. The Staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set for the 10 CFR Part 51.22(c)(9). Pursuant to 10 CFR Part 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

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6.0 CONCLUSION

We have' concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the will not be endangered by operation in the proposed manner, (2) public such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: Herb Conrad Dated: June 5, 1989

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