ML18191A257

Letter on the Schedule for Revision to Plant Design Assessment Report of WPPSS Nuclear Project No. 2

ML18191A257
Person / Time
Site:	Columbia
Issue date:	02/15/1977
From:	Renberger D Washington Public Power Supply System
To:	Varga S Office of Nuclear Reactor Regulation
References
G02-77-59
Download: ML18191A257 (10)

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Addi info & schedule for revision to plant design assessment report..................

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Ilatory Docket File Washington Public Power Supply System A JOINT OPERATING AGENCY P. O. BOX 96 INGTON WAY RICHI.ANO, WASHINOTON 99352 PHONC(509) 946 1611 Docket No. 50-397 oo--:119 USMC II,8 18Vv IO February,15, 1977 G02-77-59

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Attention:

Mr. S.

A. Varga

'4 Director of Nuclear Rea lation U. S. Nuclear Regulatory Commission Washington, D.

C.

20555

Subject:

WPPSS NUCLEAR PROJECT NO.

2 SCHEDULE FOR REVISION'TO PLANT DESIGN ASSESSMENT REPORT

References:

1) 2)

3)

Gentlemen:

Letter D. L. Renberger to S.

A. Varga, entitled "Mark II Containment Dynamic Forcing Functions Information Report,"

dated September 28, 1976 (G02-76-424).

Letter D. L. Renberger to W.

R. Butler, entitled "Mark II Containment - Design Evaluation," dated March 18, 1976 (G02-76-98).

Letter J.

F. guirk to W.

R. Butler, entitled "Mark II Phase I - 4T Tests Application Memorandum," dated June 14, 1976.

In Reference 1,

we indicated that we planned to submit a revision of the original WPPSS Nuclear Project No.

2 (WNP-2) Plant Desi n Assessment for Safet -Relief Valve and Loss-of-Coolant Loads transmitted by Reference 2) reflecting the WNP-2 capability to accommodate the loading given in Revision 2 of the Mark II Owners Dynamic Forcing Function Information Report (DFFR Rev. 2).

Because of the ongoing Mark II Owner programs to better define the chugging phenomena and quencher safety-relief valve discharge device loading characteristics, we believe it would be more meaningful to submit a revised assessment report incorporating results of these investigations when they become available.

These load definition investigations will limit the uncertainties involved to allow more meaningful information to be used in the plant evaluation.

Accordingly, the potential for unnecessary plant modifications decreases with the availability of more meaningful loading information and, therefore, we are now scheduling our revision to the assessment report to be completed not later than December 1977.

Since the WNP-2 schedule does not dictate that a revised design assessment report be submitted immediately, as is the case for the lead Mar k II plants, we believe it would be more

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Director of Nuclear Reactor Regulation Page 2

worthwhile to concentrate our efforts (also NRC efforts) on the chugging and quencher SRV load definition work rather than providing an interim revision to the WNP-2 assessment report at this time.

The peak value of the forcing function (commonly referred to as "bubble pressure") f'r SRV discharge which'as been used in the WNP-2 assessment has been determined using the methods presented in the Revision 2 of DFFR.

Based on the WNP-2 SRV line air volumes, the revised quencher model results in a reduction of bubble pressure to about 751 of the magnitude used 'in the original WNP-2 design assessment report.

We have also found that significant differences (factors of two or more differences) can be obtained in structural analysis results depending on the frequency assumptions which are made.

We are currently determining a basis for selecting discrete quencher frequencies for application in the WNP-2 plant assessment.

Similarly, GE and the Mark II owners group are attempting to provide better definition of the chugging phenomena.

As in the case of SRV discharge, the frequency character of the forcing function will have an important influence on the calculated structural response.

Information presented in Exhibit I of this letter concerning chugging is based on the current forcing function described in Reference 3.

We anticipate this chugging definition will be improved considerably with the Mark II owners undertaking an analytical evaluation of the available exper i-mental chugging data.

In view of our decision to submit a revision of the design assessment report for your review at a later date, we want to keep you informed of our plans for installation of containment stiffener r ings which were identified to be required in our original analysis.

We have had Burns and Roe, our Architect-Engineer, perform studies of the capability of the containment vessel using a variety of assumed loading conditions and a variety of stiffener designs.

The stiffener design selected for installation in WNP-2 is shown in Exhibit 2.

The conservatism associated with this design is illustrated in Exhibit I which compares the vessel capability with the current load predictions.

The structural evaluation methods were presented in the original WNP-2 design assess-ment report.

The stiffening design provides a capacity considerably in excess of the currently estimated loads.

This large degree of conservatism is not necessarily a reflection of our estimate of the true uncertainties involved but rather represents the fact that cost studies of a variety of stiffener arrangements demonstrated that the added expense was minimal for a stiffener design of the type chosen as compared to other possibilities with less load carrying capabilities.

Di,rector of, Nuclear Reactor Regulation Page 3

WPPSS has been examining the impact of the identified required containment modification and has determined the time required to implement the design modification. In order to meet the project schedule'for fuel load, the suppression pool must be available prior to June 1979 for system startup testing.

It is estimated that the preparatory'ork and installation of the containment retrofit will r equire two years to perform.

Consequently, 'initiation of activities associated with implementation of the retrofit has'een started.

In view of the fact we are proceeding with construction modifications in the-absence of final conclusions of the dynamic, loading uncertainties, WPPSS retained an independent consultant to assess the Burns and Roe recommended course of action.

The consultant (Nuclear Technology Incorporated-NUTECH) has been intimately involved in containment vessel design and analysis of several other boiling water reactor projects including the Mark I resolution of the hydrodynamic loading phenomena.

The consultant's general conclusion was that the reinforcement of the containment vessel was necessary and that the approach being taken was conservative.

If you have any questions regarding this matter, please do not hesitate to contact us.

Very truly yours, DLR:GLG:df D. L.

RENBERGER, Assistant Director, Generation 8 Technology cc:

D.

C. Baker BSR J. J.

Byrnes -

BIER D.

Roe - BPA J. J. Verderber -

B&R

A.

SYMMETRIC LOADING CONDITION EXHIBIT 1 CONTAINMENT VESSEL LOAD VERSUS CAPABILITY SRV Firing Condition E UIVALENT STATIC LOAD DUE TO:

LOCA 1)

SRV Pressure Discharge Chugging LOCA + SRV

+ Chugging S1 S1 S1 S1 CAPABILITY E UIVALENT STATIC LOAD Existing Wetwell Without Stiffeners S1 Wetwell With Stiffeners S1 ADS ACTUATION ACTUATION OF ALL VALVES 38 38 25 36 20 66 45 45.

124 124 B.

NON-SYMMETRIC LOADING CONDITION ACTUATION OF 25 TWO ADJACENT SRV's 25 32 66 Not Calculated 100 NOTES:

(1)

LOCA pressure represents pressurization of the wetwell as a result of LOCA.

(2)

WNP-2 average SRV line air volume is about 72 FT

, with resulting average peak positive quencher bubble pressure of about 15.6 psi based on )FFR, Revision 2.

Similarly for ADS actuation average air volume is about 72 FT and average peak positive quencher bubble pressure is about 14.9 psi.

The increase from these pressures to those shown in the table above is due to dynamic amplification.

(3)

Chugging equivalent static load is based on application memorandum specification of chugging forcing function (15 psi).

(4)

Maximum non-symmetric pressures for SRV and chugging loads are conservatively assumed to occur at the same location.

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