ML20141K325
| ML20141K325 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 01/15/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20141K317 | List: |
| References | |
| NUDOCS 8601220449 | |
| Download: ML20141K325 (5) | |
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4 UNITED STATES p,
NUCLEAR REGULATORY COMMISSION 1
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\\*****/ SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 91 TO FACILITY OPERATING LICENSE N0. DPR-3 YANKEE ATOMIC ELECTRIC CCMPANY YANKEE NUCLEAR POWER STATION DOCKET NO. 50-29
1.0 INTRODUCTION
By letter dated July 19, 1985, the Yankee Atomic Electric Company (YAEC or the licensee), pursuant to 10 CFR 50.90, applied to modify its opera-tional license, DPR-3, by incorporating revisions into the Yankee Nuclear Power Station Technical Specifications (TS).
Item 1 therein proposes to realign certain Main Coolant System structural integrity surveillance require-ments for the control rod shroud tube (CRST) assembly. The licensee pro-poses that the requirements stipulated in the Safety Evaluation Report (SER) for Proposed Change 106 related to the CRST assembly be changed to those requirements stated in Section XI of the ASME Boiler and Pressure Vessel Code (Section XI).
Item 2 proposes to eliminate further surveillance in-spections of the pressurizer cladding.
The changes are to be accomplished by deleting Technical Specification (TS) requirements for in situ surveillance inspection of the CRST assembly in-dicated in 4.4.9.2 and for pressurizer applied liner cladding surveillance inspection indicated in 4.4.9.3; and utilizing the standard TS requirement stated in 4.0.5 which is written to meet the minimum ASME Section (SC)
XI requirements for the shroud tube inspection. Further surveillance inspection of cracks in the pressurizer applied liner cladding would be eliminated.
2.0 DISCUSSION AND EVALUATION 2.1, Shroud Tube Inspections The original control rod shroud tube (CRST) assembly failed following )
12 years of plant operation. Bolts attaching the CRST's (individually to the bottom plate of the reactor pressure vessel (RPV) lower core support plate (CSP) became loosened and/or sheared off which resulted in the shroud tube losing perpendicularity with the CSP which adversely affected control rod drop action. Redesign of the CRST assembly in-cluded a simplified four piece unitized assembly bolted from above (rather than from below) with the bolts extending through both the upper and lower plates of the lower CSP. The redesign also employed a recessed fit design to maintain shroud tube perpendicularity even in the case of bolt loosening. The redesign was approved in the Safety Evaluation Report (SER) written for proposed change (PC) 106. The A ranuired that the licensee perform a program of in situ TV optics 9601220449 860115 DR ADOCK O 9
' visual surveillance inspection of the shroud tube attachment at each refueling outage. This inspection was included in 4.4.9.2 of the TS in 1972. Licensee inspections, conducted at every refueling outage since the design change, to meet the T.S. 4.4.9.2 requirement consistently verified that the design modification successfully maintained CRST perpendicularity as reported in IE Inspection Report 50-029/85-19.
The proposed TS change would substitute less frequent (40 month) in situ underwater examination and more effective (120 month) exam-inations with the assembly removed from the RPV for the current in situ examination conducted at each refueling outage.
Based on this evaluation, the staff concludes that performing in-spections for the control rod shroud assemblies in accordance with Section XI can be accomplished using T.S. 4.0.5, and deletion of T.S.
4.4.9.2, is therefore, acceptable.
2.2 Pressurizer Internals Inspections The pressurizer was fabricated with an applied 304 stainless steel liner cladding where the liner was attached to the pressurizer pres-sure boundary wall by resistance spot welding. This method was used by B&W for early nuclear prototype vessels and was discontinued early in the commercial nuclear power plant business in favor of weld deposited cladding as a result of welding metallurgical problems associated with the spot welds, and due to the potential for intergranular stress corrosion cracking of the cladding in primary oxygenated water.
Early visual inspections conducted by the licensee in the mid 1960's, 1970 and 1974 identified cracking of the cladding.
Upon the institution of SC XI ISI examination programs between 1970 and 1974, the cladding has been examined by surveillance inspection methods described in the licensee TS 4.4.9.3 at every (18 month) refueling outage. The purpose of this in situ TV optics visual sur-veillance inspection was to determine if any further cracking occurred or if existing cracking propagated. The licensee evaluation of the results of periodic TV optics visual surveillance inspections indicates there have been no significant changes in the cladding cracking since inception of the inspection program. The licensee also indicated that the ultrasonic inspection (UT) conducted in 1974 demonstrated that the i
surface cracking of the cladding did not propagate into the pressure boundary wall of the pressurizer. A staff review of the applicable ISI documentation related to the licensee's inspection of the cladding cracking since 1970 is reported in IE Inspection Report 50-029/85-19.
This inspection consisted of an examination of the 1974 UT report, observation of underwater TV optics video tapes representing both early examinations and the most current examination, and review of related
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documentation. The staff examination results concurred with the licensee's evaluation that cracking is not propagating in the cladding or extending into the pressure boundary. The staff review examined the changes in SC XI, IWB-2500 where the Category B-I-2 examination requirenient for internal surface cladding cracking was eliminated in 1976. Discussions with NRR MTEB personnel indicated that the elimination of the requirement for examination of weld deposited cladding cracking was based on the engineering acknowledgement that weld deposited cladding cracking does not propagate into the pressure bounda ry. The subject cladding by its nature as an applied liner is more resistant to propagation of cracking then weld deposited cladding.
The lack of propagation in applied liner cladding has been confirmed by examination of these liners by nondestructive and destructive methods in nuclear prototype vessels.
The licensee's observations in the cladding cracking are consistent with known experience and elimination of further examination is in accordance with current SC XI requirements.
The staff concludes that elimination of further pressurized applied liner cladding examination (as defined in TS 4.4.9.3) is an accept-able change to the TS.
3.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 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 occupation 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 forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assess-ment need be prepared in connection with the issuance of this amendment.
4.0 CONCLUSION
We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; and (2) such activities will be conducted in compliance with the Con 11ssion's regulations l
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and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
5.0 ACKNOWLEDGEMENT This Safety Evaluation has been prepared by S. D. Reynolds, Jr., Division of Reactor Safety, Region I.
DATE:
January 15, 1986 4
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