ML19296C011

Safety Evaluation Supporting Amend 54 to License DPR-37

ML19296C011
Person / Time
Site:	Surry
Issue date:	12/20/1979
From:	Office of Nuclear Reactor Regulation
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ML19296C003	List: ML19296C002 ML19296C011 ML19296C016
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NUDOCS 8002250052
Download: ML19296C011 (7)

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UNITED STATES 8 '?

NUCLEAR REGULATORY COMMISSION j j, r4 E

WASHINGTON, D. C. 20555

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j SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACT 0_R REGULATION RELATED TO AMENDMENT NO. 54 TO FACILITY OPERATING LICENSE NO. DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION, UNIT NO. 2 DOCKET NO. 50-281 Introduction By letter dated December 30, 1976, as supplemented May 24, 1979, Virginia Electric and Power Company (the licensee) requested changes to the Technical Specifications appended to Facility Operating License Nos.

DPR-32 and DPR-37 for Surry Power Station, Unit Nos.1 and 2.

Discussion In August 1974, we requestad that the licensee submit proposed Technical Specification changes that would establish requirements for a program of steam generator tube inspection.

To provide guidance in developing an inspection program at that time, the licensee was to refer to Regulatory Guide 1.83, " Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes," dated June 1974 (R.G.

1.83).

The licensee submitted a program for Surry on November 18, 1974. However, we made a decision to delay requiring Technical Specification incorporation of the program at that time because of a need to revise R.G.1.83 to reflect developments in the state-of-the-art of steam generator tube inspection techniques and to more directly take into account the inspection experience that was being gained at operating plants.

In making that decision we took into account the industry wide practice which already included voluntary inspection of steam generator tubes that in many respects, was comparable to inspections that R.G.1.83 specified.

Revision I to R.G.1.83 was issued after receiving comments from the industry. We are now taking steps to require incorporation of steam generator tube inservice inspections into the Technical Specifica-tions. By letter dated December 30, 1976 as supplemented May 24,1979, the licensee proposed Technical Specifications which reflect the provisions of R.G.1.83, Revision 1, with exceptions as discussed with the NRC staff.

The Technical Specifications proposed for the Surry steam generator tube inspections are, therefore, in general agreement with R.G.1.83, Rev.1, o%

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. dated July 1975, but deviate in those areas where we have determined that the overall inspection program would be enhanced over that covered in R.G. 1.83, Rev. 1.

In addition to the proposed changes to the Technical Specifications to implement steam generator inservice inspection, the activity requirements of Appendix A-1 to the Technical Specification are being incorporated into the text of the Technical Specifications with some changes as discussed with the licensee.

The proposed changes to the Technical Specifications discussed in this Safety Evaluation Report are intended to be applicable to both Surry 1 and Surry 2.

However, these changes will not be applied to Surry 1 until startup after steam generatcr repair on Surry 1.

In the interim, inspection requirements on tnat facility are governed by existing license conditions.

Evaluation I.

Surveillance Requirements for Steam Generator Tubes Structures, syste,s, and components, inportant to safety of a nuclear power plant are cesipned, f abricated, constructed, anc tested so as to provice reasonable assurance that the facil ty can be operated without undue risk to the health and safety of tne public. To continuously naintain such assurance, General Design Criterion 32 recuires that components which are cart of the reactor coolant pressure boundary (RCPB) be designed to permit perindic inspection and testinp of important areas and features to assess their structural and leaktight integrity. The stean generator tubing is part of the R:PE and is an important part of a major barrier against fission product release to the environment.

It also acts as a barrier against steam release to the Contalnnent in the event of a Loss of Coolant accident (LOCA).

For this reason, a orogran of perioaic inservice inspection is being established to assure the continued integrity of the stea cenerator tunes over the service life of the ol3nt.

Generally, the major elements of the steam generator tube inservice inspection program for Surry consist of specified:

(a) sample selection, (D) examination methods, (c) inspection intervals, (d) acceptance criteria, and (e) reporting recuirements.

Each of these major elements of the program is separately evaluated below.

(a) Sample Selection The proposed sampling program is patterned af ter R.G. 1.83, Rev. 1, except for tnose deviations that we have determinec will improve tne program and thereby reduce the potential radiation exposures especially for personnel that must perform the inspections. The samplinc procedures are contained in

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. Table 4.19-2 of the proposed Technical Specifications.

The principal deviations from R.G.1.83, Rev. 1, supple,entary sampling requirements are evaluated below.

(i)

Regulatory Position C.5.a, " Supplementary Sampling

'equirements,' recomnends tnat if the eddy current inspection results during an inservice inspection indicate any tubes with previously undetected imperfections of 20% or greater depth, additional steam generators, if any, should be insoected. In other words, because of a single tube in one steam generator with a previously undetected imperfection of 202, or greater, depth, but still well below the plugging limit, all steam generators in the unit would be inspected.

This would be unreasonably severe and would increase unnecessarily the radiation exposures of inspection personnel.

The supplenentary sampling recuirements, as modified, would still require insoection of additional steam generators where significant, but only if the inspection results of the particular steam generator f all in category C-3 which is defined in Specification 4.19.C as "more than 10% of the total tuces inscected are degraded tubos or more than I of the inspected tubes are def ec tive.

sy thus ~191m121cg t:e ins,ection cf otner steam generatcrs, the er?ossre to rers n,1 ca ce kept Irw as is reasonably e:" evar'e.

(ii) Revision I of R.G. l.83 recommenas additional tube inspections in a stean generator if the inspection of a sample of tubes results in more than lot of the tubes in tnat initial sanple having detectable wall penetration of greater than 20% or if one or more tubes in the sample have an indication in excess of the plugging limit.

The first additional inspection recommended by the guide would be the inspection of an additional 3% of the tubes in that steam generator concentrating on those areas wnere imperfections have been found.

If 10% of tnese additionally inspected tubes fail to meet the criteria acclied to the initial inspection sample, a tnird sample consisting of at least 6% of the tubes in that steam generator in the area of the imperfections would be expected.

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_4 For addition 31 inspections, if required af ter the initial inspections, the program set forth in the Surry prop ^ sed Technical Specifications would require that all tubes in the affected steam generator be inspected and, further, tnat ti. ice the nunter of tubes initially inspected in the af fected steam generator De inspected in each of the other steam generators. Again, if more than lot of tne tubes inspected in any steam generator have indications of wall tninning of greater tnan 20s or more tnan le of the tubes are defective a thira inspection is require

• of all tubes in each steam generator with sucn indications.

The primary purpose of the additional inspections is to confirm the initial insDection results and to ensure stean generator integri ty.

By recuiring tnat all tubes in the affected steam generator be inspected, and by recuiring tnat the other steam generators De inspected wnen problems are detected, the Surry program reDresents an improvenent to R.G.1.63, Rev. 1.

Eased on tne considerations discussed aDCve, we have concluded that the sanple selection scheme proposed of tne licensee is accept 3 Die.

(b) Examination fiethod Tne proposed exaraination methods, as modified by the NPC Staf f and concurred in by the licensee, include nondestructive examination Dy eday current testing.

The specified methods are capable of locating and identifying stress corrosion cracks and tube wall thinning from chemical wastage, mechanical damage or other causes. Based on our review of these methocs, and experience gainea using these methocs by the incustry, we have concluded that the examination methods are acceptable.

(c)

Inspection Intervals The proposed in5Dection intervals are compatible with *_ hose recommendec in R.G.

l.e3, Rev.

1-, and thus, are acceptable.

(c) Acceotance Criteria The principal paraneter used to deternine whether any one ste3n generator tuce is acceptable for continued service is the measured imperfection deptn. A tube pluccina linit has been establisned and cefined in tre Technical Specifications as the interfection depth beyond which tne tuDe nust be ronovec from service. The plugging limit is 40% degradation of the nominal tube wall-thickness.

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. The plugging linit is bas?d on (I) the ninimum tute wall thickness needed to maintain steam generator tube irtegrity during the lie.itina stress loadincs associated with a LOCA co Dined with a Sate Shutccwn Earthouake (SSE ), ano (2) an o-era tional alloaarce to account for the time interval cet een i,nspections.

Based on other evaluations made by the NRC staff, analyses performed by West gh:use cn ste3-generator tune designs similar to i

the Surry tube Jesign, and Regulatory Guide 1.121, Bases for Flucgina Cegradea W.A Steam annerator :;?es,

,.e concluce that the proposed 40% degradation plugging limit will ensure the required factors of safety of three against tube rupture under normal operating conditions and will provide a margin of safety consistent with Section III of the ASME Boiler and Pressure Vessel Code under postulated accident ccaditions including LOCA & SSE.

Furthermore, the proposed plugging limit includes a sufficient thickness degradation allowance to cocnensate for oossible continued degradation between inscryice ins? actions.

Tieraf:ra, Na find a t tha ?rocosed CC1 (agr-drtic.' ni upcing lini. is acce?tabl e.

(e) Reportino of Inspection Results Regulatory Position C.7.d of R.G.1.83, states tnat a licensee should report to the Commission, for resolution and approval, proposed remedial action if the inspection results excees the lirits specified in tne Guide.

It also states that adcitional sampling anc more frequent inspection may be required. The proposed Tecnnical Specifications, as modified by tne NRC staff and concurred in by the licensee, clearly specify additional inspections tne licensee must perform for those inspection results that fall in Categories C-1 anc C-2.

Inmediate reportinc of these results would not be required.

Immeciate reportin;, as indicated by Table TS 4.19-2 would be requirec only if the inspection results reachec the relatively greater level of tube decradation and defect founc in Category C-3.

We conclude that the above describec reporting recuirements, as proposed by the licensee and noo1fied Dy us, are reasonacle and will f acilitate reporting of pertinent inforuation witnout unnecessarily increasing plant downtime, and thus constitute an accepte:le alternative metnod for ccmplying wi th the Commission's re;aiations.

!n sum 3ry, we have concluded tnat tre p oposed stean generator tuDe inservice insuection prt; ram will provide added assurancc of the continvec integrity cf tne steam generator tt Des, and thus is acceptaDie.

Suoplemental Testinony of Ja es P.

Enir"t Dcf:re ne Ato-4: 53'et, and Licensing Arpeal board

.7-The Matter cf N ;r:rerr States P>ar Cotrany, Docket Nos. 40-2:2/3co.

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, 2.

Reactor Coolant to Steam Generator Secondary Side Leak Rate Limit (a)

The existing license condition of 0.3 gpm specifies a reactor coolant leak rate limit. The proposed change would specify a reactor coolant leakage rate limit of 500 GPD (about 0.35 GPM) from any one steam generator, and would require that the leakage be reduced to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or that the plant be placed in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> if the leak rate is exceeded.

The limit of 0.3 gpm was established to assure an early shutdown for inspection upon detection of very small leaks taking into account the state of degradation of the plants steam generators prior to shutdown to accomplish the steam generator repair program in which the entire lower assemblies of the steam generators, including all tubes, were replaced.

These new lower assemblies are now in plam in Unit 2 and constitute essentially new condition stea.

ierators insofar as the primary-to-secondary pressure boundary is concerned.

The standara f6RC Technical Specifications for Westinghouse plants all use this 500 GPD limit for new steem generators.

We have, therefore, concluded that a 500 GPD leakage rate limit for each Surry 2 steam generator is acceptable at this time.

The proposal to move the steam generator inservice inspection requirement from the body of the license to the Appendix A Technical Specifications should be accompanied by a similar reloca-tion of the related reactor coolant and secondary coolant activity limits incorporated into the Technical Specifications.

These limits should be essentially the same as they are now with the exception of changing the total specific activity limit of the reactor coolant to 100/E from 41/E for nuclides with hal f-lives of more than 15 minutes. These changes are necessary to be consistent with the specifications used for other cases and with the standard NRC Technical Specifications applied to other Westinghouse reactors. The licensee agrees with these changes.

These limitations on reactor coclant specific activity will ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steady state primary-to-secondary steam generator reactor coolant leakage rate of 1.0 GPM. The values for the limits on specific activity represent interim limits based upon a conservative application as a parametric evaluation by the NRC of typical sites.

The NRC is finalizing site specific

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, criteria which will be used as the basis for this reevaluation of the specific activity limits of the Surry site.

This reevaluation may result in higher limits for the Surry site.

E nvi ronmental Consideration We nave deterninea that the amendment does not authorize a Change in types or total anounts nor an increase in power level and will effluent not result in any significant environnental impact.

Having nace this dete-mination, we have further concluded tnat the amendnent involves action which is insignificant from the stanapoint of environmental aninpact and pursuant to 10 CFR %51.5(d)(4), that an environ ental impact statenent or negative declaration and environmental impact appraisal neec nct be prepared in connection with the issuance of this anenament.

Conclusion based on the considerations discussed aDove, that'

e have concluded, cecause tne a"encrent aces not involve a significant increase in (1) the procability or consequences of accidents previousiv considerec arc does not involve a significant decrease in a safety nargin, the anend'ent does nat involve a significant hazards consiceration, (2) there is reasonaale assurance that the health ano safety of the public will

ion in tne pronosed nanner, and (3) such not be endangered by ope activities will be condu ced in compliance with the Commission's regulations and the issuance of this anendnent Will not be ini-ical to the conman defense and security or to the health and safety e f the public.

Date: Dec ember 20, 1979