ML20196C505

Informs That Attached Draft Request for Addl Info Transmitted by Facimile on 990617 to G Salamon of Pse&G. Review of RAI Would Allow Licensee to Determine & Agree Upon Schedule to Respond to RAI

ML20196C505
Person / Time
Site:	Hope Creek
Issue date:	06/21/1999
From:	Richard Ennis NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
TAC-MA4471, NUDOCS 9906240055
Download: ML20196C505 (5)

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• June 21, 1999 MEMORANDUM TO: Docket File  !

Original signed by:

FROM: Richard B. Ennis, Project Manager, Section 2 Project Directorate I

. Division cf Licensing Project Management '

Office of Nuclear Reactor Regulation

SUBJECT:

HOPE CREEK GENERATING STATION, DRAFT REQUEST FOR ADDITIONAL INFORMATION, INCREASE OF ALLOWABLE MAIN STEAM ISOLATION VALVE (MSIV) LEAK RATE AND DELETION OF MSIV SEALING SYSTEM (TAC MA4471)

The attached draft request for additional information (RAI) was transmitted by facsimile on ' June 17,1999 to Mr. Gabe Salamon of Public Service Electric & Gas Company (PSE&G).

Review of the RAI would allow the licensee to determine and agree upon a schedule to respond to the RAl. This memorandum and the attachment do not convey a formal request for information or represent an NRC staff position.

Docket No. 50 354

Attachment:

Draft RAI S C Rf CENTS CDPT DISTRIBUTION Docket File -

PUBLIC JClifford.

REnnis OFFICE ' PDI-2/PM l NAME- REnnis Y E l [/[!00 4 DATE 0 /M /99 l OFFICIAL RECORD COPY DOCUMENT NAME: G:\PDI-2\ Hope Creek \mema4471.wpd 9906240055 990621 PDR P

ADOCK 05000354 pg

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j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001

%.....p June 21, 1999 MEMORANDUM TO: Docket File FROM: Richard B. Ennis, Project Manager, Section 2 Project Directorate l Division of Licensing Project Management Office of Nuclear Reactor Regulation

SUBJECT:

HOPE CREEK GENERATING STATION, DRAFT REQUEST FOR ADDITIONAL INFORMA TION, INCREASE OF ALLOWABLE MAIN STEAM ISOLATION VALVE (MbiV) LE.4K RATE AND DELETION OF MSIV SEALING SYS1 EM (TAC MA4471)

The attached draft request for additional information (RAI) was transmitted by facsimile l on June 21,1999 to Mr. James Priest of Public Service Electric & Gas Company (PSE&G).

i Review of the RAI would allow the licensee to determine and agree upon a schedule to respond to the RAl. This memorandum and the attachment do not convey a formal request for information or represent an NRC staff position.

l l

Docket No. 50-354

Attachment:

Draft RAI

REQUEST FOR ADDITIONAL INFORMATION INCREASE OF ALLOWABLE MAIN STEAM ISOLATION VALVE (MSIV) LEAK RATE AND DELETION OF MSIV SEALING SYSTEM HOPE CREEK GENERATING STATION (TAC NO. MA4471)

References:

1. Letter from E. C. Simpson (PSE&G) Company to the Document Control i Desk (NRC), " Request for Change to Technical Specifications, increase of Allowable Main Steam Isolation Valve (MSIV) Leak Rate and Deletion of MSIV Sealing System," dated December 28,1998 )

2. Letter from F. M. Akstulewicz (NRC) to T. A. Green (BWROG)," Safety Evaluation of GE Topical Report, NEDC-31858P, Revision 2, 'BWROG Report for increasing MSIV Leakage Limits and Elimination of Leakage Control Systems,' September 1993," dated March 3,1999

1. Provide a detailed description of the altemate leakage treatment (ALT) pathway and the basis for its functional reliability, commensurate with its intended safety-related function.

Also, provide a descriptic' ->f the maintenance and test!ng program for the active components (such as valus) in the ALT pathway.

2. Clarify whether all pipe support anchorages in the ALT pathway have been seismically analyzed. if not; identify the pipe support anchorages that were not analyzed, and provide justification for the statement, made in Section 4.4 of Attachment 4 to Reference 1, that "all support anchorages have adequate capacities," without having all pipe support anchorages analyzed. ';; , , , _

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3. Discuss whether;the loading at the pipe support anchoragesp, as generated from the seismic analysisbf pipjng systems. If rgt, describe the meth%used.

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4. Describe the r}$thoisfi$ critdrid uhed t2 Otita 6ibe capacity $f Nipe support anchorage.

5. In Section 4.4 of Attachment 4 to Reference 1, you stated that pipe supports for the non-seismically designed portion of the ALT pathway have been evaluated using the Conservative Deterministic Failure Margin (CDFM) methodolog; from EPRI Report NP-6041. This methodology has not been approved by the NRC, as discussed in Reference 2. Therefore, a plant specific seismic evaluation for representative supports and anchorages associated with the non-seismically designed portion of the AtT patnway should be performed. The evaluation shou!d be performed using the plant 1 licensing basis methodology, or other methods acceptable to the staff. From this plant  !

specific evaluation, provide a comparison of the resulting support loads to their capacities and the associated safety margins.

l ATTACHMENT

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6. In relation to item (6) above, provide calculations for a typical pipe support anchorage that serve to illustrate the process of demonstrating the seismic adequacy of the support anchorage. ,

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7. Provide a bounding seismic analysis for the ALT pathway, subject to all the pertinent i design loading. Discuss the basis for the selection of the analyzed portion of the drain i line piping for the bounding analysis.

8. Provide your approved plant walkdown verification procedure for Hope Creek's ALT pathway. .

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9. On page 3-1 of Attachment 4 to Reference 1 the high pressure condenser at Hope Creek is compared to similar condensers at Moss Landing Units 6 & 7 and Ormond Beach Units 1 & 2. The first sentence of the third paragraph on page 3-1 of Attachment 4 states, "_In summary, the condenser design and anchorage are similar to those at facilities in the earthquake experience database that have experienced earthquakes in excess of the Hope Creek design basis SSE (See Figure 4-1)." The Moss Landing response spectrum shown on Figure 4-1 of Attachment 4 is not the same as the spectrum that has been previously accepted by the staff. The response spectrum for Moss Landing, estimated from ground motion from the 1989 M6.9 Loma Prieta earthquake, that has been accepte oy the staff was developed by Pacific Gas &

Electric (PG&E)qFurthe.rmore, the Ormond Beach Po,wer Plant response spectrum, used because the condenser at Ormond Beach Power; Plant Is similar to the Hope s

Creek condense (Is not' plotted,on Eigure,(4-1. Provide a ses igte p 1 the Ormond Beach Powerflangesponse spectrum and the borrect Moss Landing  ;

response spectrum. $ p gh $j is 10.

b M d 8#w M M 4 On page 4-4 of %decti$ent 4;toI Referede ddfirsthragrkhh the section entitled

" Comparison 6f HopdCfeek D6 sign SSE ~I Spebtiswith ths Earthh6ake Database Plants" states, "The Hope Creek design basis SSE ground response spectrum was compared with the ground motion spectra at several database power plant sites in the attached Figure 4-1. From a review of Figure 4-1, the database spectra is seen to ylgnificantly envelope the Hope Creek spectrum over the entire frequency range of interest."

Provide the frequency range of interest referred to above since the Valley Steam, NRC-approved Moss Landing, and the Ormond Beach spectra (see Reference 2) do not envelope the Hope Creek SSE design spectrum over all frequencies.

11. Figure 4-1 of Attachment 4 to Reference 1 shows zero period acceleration (ZPA) values for 4 facility experience database ground motions. It is the staff position that although peak ground acceleration has been used in the past to characterize earthquake strong ground motion, this single parameter does not have a good correlation with earthquake damage.L A much better correlation of ground motion damage potentialis the ground response spectrum which demonstrates the maximum amplitude of the ground motion as a function of the natural frequency. It is the NRC position that the appropriate characterization of the ground motion at a facility, to be u:ad to verify the adequacy of equipment similar to that in nuclear power plants, is the response spectra developed

3-from the ground motion recorded at or near a facility.

i The staff has accepted the Humboldt Bay response spectra from the 1975 Femdale I earthquake and the 1992 Petrolia earthquake as well as the Glendale response i spectrum from the 1971 San Femando earthquake as part of the earthquake database ground motion (Reference 2). If equipment from the Humboldt Bay Nuclear Power Plant or Glendale Power Plant is used to qualify equipment at Hope Creek, then provide a separate plot showing the Hope Creek SSE design spectrum and the entire 1975 and 1992 Humboldt Bay response spectra and the entire Glendale response spectrum.

12. In Table 1," Dose Comparisons," of Attachment 1 to Reference 1, you have provided control room operator doses for a postulated design basis accident for 30 days. Provide the unfiltered control room air infiltration rate assumed in the control room operator dose calculations and its bases. State if you have performed any control room unfiltered air inteakage test.

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