ML20196J442

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Forwards Request for Addl Info Re Increase of Allowable Main Steam Isolation Valve (MSIV) Leak Rate & Deletion of MSIV Sealing Sys for Plant
ML20196J442
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 07/01/1999
From: Richard Ennis
NRC (Affiliation Not Assigned)
To: Keiser H
Public Service Enterprise Group
References
TAC-MA4471, NUDOCS 9907070283
Download: ML20196J442 (6)


Text

Mr. Hirold W. K:iser July 1,1999 Chi:f Nucl:ar Offic:r & Pr:sid:nt -

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'Nucl:;r Busin:ss Unit Public Service Electric & Gas Company Post Office Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

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)

Dear Mr. Keiser:

In a letter dated December 28,1998, Public Service Electric and Gas Company (PSE&G) submitted an application for an amendment to the Technical Specifications (TSs) for Hope Creek Generating Station. The proposed amendment would revise the TSs to permit an increase in the allowable leak rate for the main steam isolation valves (MSIVs) and to delete the MSIV Sealing System. The main steam drain lines and the main condenser would be utilized as an alternate MSIV leakage treatment method.

The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure. We request that the additionalinformation be provided within 60 days of receipt of this letter. The 60-day response timeframe was discussed with Mr. James Priest of your staff on June 24,1999. If circumstances result in the need to revise your response date, or if you have any questions, please contact me at (301) 415-1420.

Sincerely, original signed by:

Richard B. Ennis, Project Manager, Section 2 Project Directorate i Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-354

Enclosure:

Request for Additional Information f

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\\[.....f July 1,1999 Mr. Harold W. Keiser l

Chief Nuclear Officer & President -

Nuclear Business Unit Public Service Electric & Gas Company Post Office Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

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)

Dear Mr. Keiser:

In a letter dated Decembe.r 28,1998, Public Service Electric and Gas Company (PSE&G) submitted an application for an amendment to the Technical Specifications (TSs) for Hope Creek Generating Station. The proposed amendment would revise the TSs to permit an increase in the allowable leak rate for the main steam isolation valves (MSIVs) and to delete the MSIV Sealing System. The main steam drain lines and the main condenser would be utilized as an alternate MSIV leakage treatment method.

The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure. We request that the additional information be provided within 60 days of receipt of this letter. The 60-day response timeframe was discussed with Mr. James Priest of your staff on June 24,1999. If circumstances result in the need to revise your response date, or if you have any questions, please contact me at (301) 415-1420.

Sincerely, Richard B. Ennis, Project Manager, Section 2 Project Directorate l Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-354

Enclosure:

Request for Additionalinformation cc w/ encl: See next page i

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' Hope Creek Generating Station cc:

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~ Jeffrie J. Keenan, Esquire Manager - Joint Generation Nuclear Business Unit - N21 Atlantic Energy P.O. Box 236 6801 Black Horse Pike Hancocks Bridge, NJ 38038 Egg Harbor Twp., NJ 08234-4130 Hope Creek Resident inspector Richard Hartung U.S. Nuclear Regulatory Commission Electric Service Evaluation Drawer 0509 Board of Regulatory Commissioners Hancocks Bridge, NJ 08038 2 Gateway Center, Tenth Floor Newark, NJ 07102 Mr. Louis Storz Sr. Vice President - Nuclear Operations Lower Alloways Creek Township Nuclear Department c/o Mary O. Henderson, Clerk l

P.O. Box 236 -

Municipal Building, P.O. Box 157 Hancocks Bridge, NJ 08038 Hancocks Bridge, NJ 08038 General Manager - Hope Creek Operations Mr. Elbert Simpson Hope Creek Generating Station Senior Vice President-P.O. Box 236 Nuclear Engineering Hancocks Bridge, NJ 08038 Nuclear Department P.O. Box 236 Director - Licensing Regulation & Fuels Hancocks Bridge, NJ 08038 Nuclear Business Unit - N21 P.O. Box 236 Hancocks Bridge, NJ 08038 Regional Administrator, Region i U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA '19406 Dr. Jill Lipoti, Asst. Director Radiation Protection Programs NJ Department of Environmental Protection and Energy CN 415 Trenton, NJ 08625-0415 1

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REQUEST FOR ADDITIONAL INFORMATION INCREASE OF ALLOWABLE MAIN STEAM ISOLATION VALVE (MSIV) LEAK RATE AND DELETION OF MSIV SEALING SYSTEM

References:

1.

Letter from E. C. Simpson (PSE&G) to the Document Control 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 alternate leakage treatment (ALT) pathway and the basis for its functional reliability, commensurate with its intended safety-related function.

Also, provide a description of the maintenance and testing program for the active components (such as valves)in the ALT pathway. (See Note 1) 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 o; Attachment 4 to Reference 1, that "all support anchorages have adequate capacities," without having all pipe support anchorages analyzed. (See Note 1) 3.

Discuss whether the loading ut the pipe support anchorages was generated from the seismic analysis of piping systems.~ if not, describe the method used. (See Note 1) 4.

Describe the method and criteria used to obtain the capacity of a pipe support anchorage. (See Note 1) 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) methodology 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 ALT pathway should be performed. The evaluation should be performed ucing the plant licensing basis methodology, or other methods acceptable to the Maff. From this plant-specific evaluation, provide a comparison of the resulting su') port loads to their capacities and the associated safety margins. (See Note 1)

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In relation to item (5) above, provide calculations for a typical pipe support anchorage that serve to illustrate the process of demonstrating the seismic adequacy of the support anchorage. (See Note 1) 7.

Provide a bounding seismic analysis for the ALT pathway, subject to all the pertinent design loading. Discuss the basis for the selection of the analyzed portion of the drain line piping for the bounding analysis. (See Note 1) 8.

Provide your approved plant walkdown verification procedure for Hope Creek's ALT pathway. (See Note 1) 9.

On page 3-1 of Attachment 4 to Reference 1 the high-pressure condenser at Hope Creek is compared to similar condsnsers 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 Safe Shutdown Earthquake (SSE) (See Figure 1

4-1)." The Moss Landing response spectrum shown on Figure 4-1 of Attachme'it 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 accepted by the staff was developed by Pacific Gas & Electric (PG&E). Furthermore, the Ormond Beach Power Plant response spectrum, used because the condenser at Ormond Beach Power Plant is similar to the Hope Creek condenser, is not plotted on Figure 4-1. Provide a separate plot of each of the database response spectra and the Hope Creek SSE design spectrum including plots of the Ormond Beach Power Plant response spectrum and the correct Moss j

Landing response spectrum. (See Note 1) 1 10.

On page 4-4 of Attachment 4 to Reference 1, the first paragraph of the section entitled

" Comparison of Hope Creek Design SSE Spectra with the Earthquake 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 l

Figure 4-1. From a review of Figure 4-1, the database spectra is seen to significantly envelope the Hope Creek spectrum over the entire frequency range of interest."

4 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 i

envelope the Hope Creek SSE design spectrum over all frequencies. (See Note 1) 11.

Figure 4-1 of Attachment 4 to Reference 1 shows zero period acceleration (ZPA) values for four facility experience database ground motions. It is the staff's 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. A much better correlation of ground motion damage potential is the ground response spectrum which demonstrates the maximum amplitude of the ground motion as a function of the natural frequency. It is the NRC's position that the appropriate characterization of the ground motion at a facility, to be used to verify the adequacy of equipment similar to that in nuclear power plants, is the response spectra developed from the ground motion recorded at or near a facility.

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k 3-i The staff has accepted the Humboldt Bay response spectra from the 1975 Ferndale earthquake and the 1992 Petrolia earthquake as well as the Glendale response spectrum from the 1971 San Fernando 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 l

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.

(See Note 1) 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 inleakage test. (See Note 2)

Notes:

1.

The regulatory basis for this question is Appendix A to 10 CFR Part 100. Specifically,Section VI of Appendix A of 10 CFR Part 100 requires that structures, systems, and components necessary to assure the capability of the plant to mitigate the consequences of accidents, which could result in exposures comparable to the guideline exposures of 10 CFR Part 100, be designed to remain functional during and after a Safe Shutdown Earthquake (SSE). Therefore, the proposed ALT pathway and main condenser are required to remain functionalif the SSE occurs. (Reference Standard Review Plan (SRP) Section 3.2.1).

2.

The regulatory basis for this question is General Design Criteric 'GDC) 19 of Appendix A of 10 CFR Part 50, as it relates to maintaining the control room in a safe, habitab!e condition under accident conditions by providing adequate protection against radiation (Reference SRP Section 6.4).

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