ML20245J037

Responds to NRC 890706 Ltr Re Violations Noted in Insp Repts 50-266/89-04 & 50-301/89-04.Corrective Actions:Evaluation of 15 Piping Supports W/Integral Welded Attachments on Local Pipe Stress Performed Utilizing Different Techniques

ML20245J037
Person / Time
Site:	Point Beach
Issue date:	08/04/1989
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	Davis A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
Shared Package
ML20245J043	List: ML20245J037 ML20245J269
References
CON-NRC-89-096, CON-NRC-89-96 IEB-79-14, VPNPD-89-429, NUDOCS 8908170440
Download: ML20245J037 (7)

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\ Electnc POWER COMPANY 231 W Michigan. PC. Box 2046. Mdwovkee. WI 53201 (414)221 2345 l I

VPNPD-89-429 NRC-89-096 August 4, 1989

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Mr. A. Bert Davis Regional Administrator, Region III U.S. NUCLEAR REGULATORY COMMISSION i 799 Roosevelt Road Glen Ellyn, Illinois 60137 Gentlemen:

1 DOCKET 50-266 AND 50-301  !

IE BULLETIN 79-14 {

POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 j i

Reference:

1) WE Letter VPNPD-89-145/NRC-89-034 to Mr. A. Bert Davis,  !

dated 3/10/89

2) WE Letter VPNPD-89-252/NRC-89-049 to Mr. A. Bert Davis, dated 4/25/89

3) WE Letter VPNPD-88-319/NRC-88-053 to Mr. A. Bert Davis, dated 6/13/88

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This letter is in response to the notice of violation transmitted by your letter dated July 6, 1989 resulting from inspection i 3

reports 50-266/89-004 and 50-301/89-004 on I.E. Bulletin 79-14 activities for Point Beach Nuclear Plant Units 1 and 2, respectively.

It also provides an update to Reference 1 and' describes actions we propose for further evaluation of our program.

i In response to Violation 1, the activities necessary to respond  !

to I.E. Bulletin 79-14 were performed under both the Wisconsin Electric and Bechtel Quality Assurance Programs. As such, pro-cedures were in place to ensure use of design bases information and that designs generated were documented and checked. The controls of the program were subject to NRC inspection during the l performance of the work and found to be adequate. i Specifically, the violation describes support calculation incon-sistency with drawings, errors of omission, or mathematical errors. In order to assure both the NRC and ourselves that the l discrepancies do not represent a programmatic weakness in the piping support qualifications, we committed to a sample program of ten Bechtel support calculations in Reference 1. Following a. ,

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. Mr. A. Bert Davis August 4, 1989 Page 2 number of conversations with Mr. Duane Danielson and Mr. Winston Liu, WE increased the sample size to fifty calculations. This I sample program is currently being performed by a third party and )

is scheduled to be completed in August 1989. We will report the '

conclusions of this review upon completion and describe any corrective actions we intend to take should they be required. I For the three specific calculations referenced in the notice of j violation we have taken the following actions:

a) Attachments 1 and 2 are the final calculation for qualification of pipe support AC-601R-2-R3G and the local stress evaluation performed pending completion of the calculation. Both of these documents have been transmitted to Mr. Danielson. Additionally, Attachment 3 provides justification for the analytical methods and assumptions used by Bechtel to qualify this support for code allowables. Attachment 3 provides the clarifications requestcd by paragraph 2d of the Inspection Reports. No further activities are required for this support.

b) The five items identified for support EB-10-A12 have been evaluated and do not invalidate the qualification of the support l as documented by Bechtel. We have corrected the error identi-fied as Item 1 and initiated drawing revisions for items 4 and 5. Item 2 has been considered acceptable by inspection.

Item 3 is not applicable since the system operates at a tempera-ture where consideration of reduced allowable stresses is not necessary.

c) The dislocation of support HB-19-HB-4B was evaluated by reanalysis of the piping system by WE in May 1988. Stresses and support loads were determined to be acceptable. These results were reported to you by Reference 3. Drawing change notices have been issued to correct the isometric drawings.

In response to violation 2, the 79-14 program criteria directed  ;

incorporation of stress intensification factors (SIF) into the j piping analysis to evaluate the effects of integral welded attachments (IWA) on local pipe stress. The inclusion of an SIF was based on engineering judgement as to whether the stress l induced was considered critical. Although not stringently proceduralized, documentation of SIFs being utilized in some 1 analyses does indicate that judgement was practiced and con-  !

sideration of effects evaluated.

We have completed the evaluations of fifteen piping supports with IWA for localized pipe stress as proposed in References 1 and 2. The evaluations were performed utilizing different  ;

l techniques than the SIF methodology.

1) Of the ten IWAs with no apparent documented consideration of an SIF, calculated local pipe stress at one IWA location exceeds B31.1 Code allowables.

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>= Mr.:A. Bert Davis August 4, 1989 Page 3

2) Of the.five IWAs with documented SIFs, calculated local-pipe stress at three locations exceeds B31.1 code allowables.

Based on these results,-we conclude that the 79-14 program criterion for consideration of local pipe stress effects from IWAs was not conservative in all cases. As stated'in Reference 2, the total number of IWA supports which required requalification by-Bechtel in the 79-14 program-was fifty-four. To date, sixteen (sample of fifteen plus AC-60lR-2-R38).have been evaluated.

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We will take the following further actions:

1) Modify and/or replace the four supports discussed above such that pipe and support stresses meet code allowables. We expect this work to be completed in January 1990.

2) Since the initial sample was biased toward " worst case" based on judgement, as described in Reference 2, we do not expect' significant overstress conditions among the remaining cases.

However, we will evaluate the remaining thirty-eight locations of IWA for local pipe stress effects. Thest evaluations are expected to be completed in December 1989.

The assumptions, methods and acceptance criteria that have been and will be used to perform theLevaluations for local pipe stress are provided and justified in Attachment 3.

An operability evaluation was performed for each of the four locations which could not be qualified for B31.1 code allowables.

These locations were qualified to operability criteria described in' Attachment 4. An example of'the use of-these criteria is provided in Attachment 5. These' operability criteria were developed originally for use at the Prairie Island Nuclear Generating Plant and are based on Appendix F of the ASME B&PV Code Section III for stress limits and load combinations. These criteria had been previously approved by the NRC for use at Prairie Island. We intend to request NRC approval of these criteria for determining operability should further discrepancies be discovered in the future.

The deportability of these operability evaluations under the pro-visions of 10 CFR Part 50, paragraphs 50.72(b)(1)(ii) and 50.73(a)(2)(ii), has been assessed. As discussed in the state-ments of consideration published with-these rules, the intent of these paragraphs is to capture those. events where the plant, including its principal safety barriers, was seriously degraded or in an unanalyzed condition. Although the inclusion of localized stress associated with integral welded attachments significantly increased the stresses, the operability evaluations for these four supports demonstrated the safety of the systems. There exists no loss of safety function in these situations and therefore, the

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.- --Mr.-A.'Bert Davis August 4,'1989 L Page 4 evaluations are not reportable under the Licensee Event Report (LER) program. Information regarding the four'overstressed IWA supports is tabulated in Table 1.

In 1988, we inspected selected piping systems to address NRC and WE concerns regarding accuracy of the walkdown information and had favorable results. A total of eleven complete and seven partial piping subsystems (anchor to anchor analyses) were evaluated. For three of these subsystems, discrepancies between the as-found piping configuration and the isometric drawing necessitated reanalysis. Mr. Liu's walkdown on May.9, 1989 of

-the auxiliary feedwater system isometrics identified additional piping supports not shown on the isometric drawings. This resulted in the walkdown and reanalyses of three additional piping subsystems. The input data was additionally corrected for modeling errors identified in Attachment 2, Item 2 of Reference.

1. These reanalyses have documented acceptability of the piping.

We are currently evaluating the impact of the reanalysis on support qualifications.

Mr. Liu's walkdown findings suggest the need for additional sampling. Therefore, we propose to perform detailed'walkdowns of fifteen additional piping subsystems. Since.a number of these additional piping subsystems are in Unit 2 containment, the.

systems are inaccessible until the Unit'2 outage beginning September 1989. It is necessary to include containment systems in this sample since the majority of systems in ' accessible areas during normal operation have already been selected for previous sampling programs, or are' located in relatively high radiation areas. After completion of these walkdowns approximately 20% of- ,

i the 79-14 program scope.will have been' evaluated. We expect to '

have all walkdown results completed and' reported to you in l December 1989.

The WE commitments described-above will-be. completed in accordance with the following schedule:  ;

1) Calculation sample program results will be provided in August  !

1989.  ;

2) Modification / replacement of the four IWAs where pipe stress exceeds allowables will be completed in January 1990.

3) Evaluation of the thirty-eight remaining IWAs will be completed in December 1989.

4) Walkdowns on fifteen additional piping subsystems will be completed in December 1989.

5) Criteria for Justification for Continued Operation (JCO) for Major "As-built" Discrepancies in Safety Related Piping

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. Mr. A. Bert Davis August 4, 1989 Page 5 1 1

Systems will be submitted for your review and concurrence in  !

August 1989. In the interim, we propose to use criteria identified in Attachment 4 for operability determinations.

We will keep you informed as we proceed through these activities.

If you have any questions, please advise.

Very truly yours, 1

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1 C. . Fay Vi'ce President Nuclear Power Attachments l Copy to Resident Inspector Document Control Desk j j

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• 4 ATTACHMENT 1 Qualification for Pipe Support AC-601R-2-R38 l

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