ML19242D657

Provides Addl Info Per 790802 Commitments,In Response to NRC 790807 Request Re Plant Startup.Schedule for Pipe Support Evaluations & Discussion of Seismic Analysis of Suppression Chamber Water Level Encl

ML19242D657
Person / Time
Site:	FitzPatrick
Issue date:	08/10/1979
From:	Early P POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	Ippolito T Office of Nuclear Reactor Regulation
References
JPN-79-47, NUDOCS 7908150676
Download: ML19242D657 (7)

Text

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POWER AUTHORITY OF THE STATE OF NEW YORK 10 CoLUMeus CIRCLE NEW YORK, N. Y. 10019 (212) 397 6200 e

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August 10, 1979 JPN-79-47 Director of Nuclear Reactor Regulation United States Nuclesr Regulatory Commission Washington, D. C. 20555 Attention: Mr. Thomas A. Ippolito, Chief Operating Reactors Branch No. 3 Division of Operating Reactors

Subject:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Supplemental Information in Support of Request for Plant Start-up

Dear Sir:

This letter provides additional information to fulfill commitments made in the Authority's letter of August 2, 1979 and in response to verbal requests made by the NRC staff on August 7, 1979.

A discussion of the conservative nature of assumptions used in our seismic analysis concerning the suppression chamber water level was requested and is furnished in Attachment 1.

The Authority's August 2, 1979 letter committed to provide, prior to start-up, a schedule to complete pipe support evalu-ations not completed at the time of start-up. This schedule, provided in Attachment 2, assigns priority to complete one of each redundant safety train expeditiously.

In analyzing the 96 piping runs, some were analyzed with the assumption that the supports were rigid. An evaluation will be initiated on or about November 1, 1979 to assess the effect of using the actual support stiffnesses for those sup-ports originally assumed to be rigid.

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U.S. Nuclear Regulatory Commission JPN-79-47 The Authority's August 2, 1979 letter committed to complete analysis and any necessary modifications for the 28 remaining pipe supports associated with the high pressure core injection system and the reactor core isolatica cooling system prior to start-up to meet DBE criteria. To date, 23 of the 28 supports have been determined to be within allowable stress limits.

A. lysis of t.*.e remaining five supports is continuing.

Very truly,yours,

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Paul J. Early Assistant Chief Engineer-Projects PJE:rz Attachments

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ATTACHMENT 1 DISCUSSION OF CONSERVATIVE NATURE OF ASSUMPTIONS CONCERNING SUPPRESSION CHM 4BER WATER LEVEL Amplified response spectra (ARS) at the suppression chamber (torus) were developed based on the dynamic characteristics of the structure and the specified ground response spectra. Studies were made to evaluate the effect of water content inside the torus on its seismic responses (FSAR 12.5.1.3). It was concluded that the chamber would be subject to higher seismic load during Design Basis Earthquake (DBE), if it was assumed " flooded". For the Operating Basis Earthquake (OBE) case, the chamber was assumed

" half full". These two cases stretched from one bound to the othar as far as the natural frequency of the system is concerned.

Since piping was analyzed both for OBE and DBE, the piping was considered to be qualified for these two bounding cases.

If the DBE is postulated to occur when the chamber is only

" half full", the seismic effect on piping may be accurately and conservatively estimated by extending the data generated for OBE.

The mathematical model is identical, and the DBE ground response spectrum is equal to the OBE multiplied by a constant factor, assuming the same damping coefficient. Consequently, the ampli-fied response spectrum and hence the stresses and loads for DBE must be equal to those of the OBE spectrum multiplied by the same constant factor for the same damping coefficient. This factor is the ratio of the maximum ground accele rations of DBE and OBE; i.e.,

0.15/0.08 = 1.875. In fact, since the damping coef ficient for DBE should be twice as much as for OBE, the actual ratio would be less than 1.5.

An ARS curve developed for the torus full of water will have a peak value co siderably higher than a curve developed for the torus only half full of water. The concern expressed by the NRC staff was that the frequency of the ARS peak will shift and a piping system resonance which would hit an ARS peak based on 2.

half full torus would miss the ARS peak based on a full torus.

Based on the above concern, the eleven cases potentially affected by the torus ARS were reexamined. The conclusions were:

a. In seven of the eleven cases, piping system natural frequencies already coincided with the peak ARS values for both the DBE and OBE cases. Therefore, use of an ARS with a different (nigher) frequency spectrum but lower maximum amplitude, should reduce the load-ings. These are cases 650, 682, 643, 693, 694, 647 and 667.

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b. In two of eleven cases, the piping natural frequencies

! do not coincide with the existing DBE peak nor would they coincide with the DBE ARS peak based on a half full torus. Therefore, stresses and leada resulting .'

d from use of revised DBE ARS curves would be lower ~

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9 cases are cases 740 and 912. . 14

c. In two cases (cases 668 and 733), the piping frequencies, '

T mode shapes and participation factors were such that y it was not absolutely clear that existing analyses, based on the torus full of water, were totally con- -

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servative. Therefore, these cases using a DBE ARS

. predicated on a half full torus were reanalyzed. It  :

was demonstrated that all pipe stress, nozzles, pene-

4 tration and supports were within code allowables. .

Based on the above, we have concluded that the analyses based

on the torus water levels specified in the FSAR are valid and - -

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conservative.

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ATTACHMENT 2 SCHEDULE TO COMPLETE REMAINING PIPE SUPPORT EVALUATIONS PRIORITY I - "A" Safety Train Systems Start - August j5, 1979 - Complete - September 15, 1979 Number of Supports Case Number to be Evaluated Residual Heat Removal 641 17 637 2 643 11 647 8 864 20 866 3 877 4 878 1 880 6 888 5 879 4 948 5 951 5 Core Spray 669 4 674 2 Service Water 863 0 865 8 874 21 875 4 876 3 881 2 900 30 901 14 691339

Number of Supports Case Number to be Evaluated Reactor Bldg. Cooling Water 872 3 873 2 Total Number of Priority I Supports to be Evaluated 184 Note: Completion of these cases will result in at least one train of all redundant safety-related systems having been evaluated.

PRIORITY II - "B" Safety Train Systcms Start - September 15, 1979 - Complete - September 22, 1979 Number of Supports Case Number to be Evaluated Residual Heat Removal 646 19 867 1 8 68 't 3 869) 870 3 871 7 Core Spray 673 0 934 5 Total ?; umber of Priority II Supports to be Evaluated 38

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PRIORITY III - Start - September 22, 1979 - Complete - Oc tober 13, 1979 Number of Supports Case Number to be Evaluaied Standby Gas Treatment 941 1 942 4 Fuel Pool Cooling 947 11 949 6 950 0 952 5 953 4 Drywell Vent & Purge 733 2 740 0 893 8 894 3 894X 6 912 .

O Standby Liquid Centrol 931 11 Fire Protection 916 23 917 9 913 12 919 21 920 15 Total Number of Priority III Supports te be Evaluated 14]

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