ML063200086

Request for Additional Information, Loss of Power Diesel Generator

ML063200086
Person / Time
Site:	Point Beach
Issue date:	11/27/2006
From:	Lyon C NRC/NRR/ADRO/DORL/LPLIII-1
To:	Koehl D Nuclear Management Co
F. Lyon LPLE X2296
References
TAC MD0936, TAC MD0937
Download: ML063200086 (6)

Text

November 27, 2006 Mr. Dennis Koehl Site Vice President Point Beach Nuclear Plant Nuclear Management Company, LLC 6610 Nuclear Road Two Rivers, WI 54241-9516

SUBJECT:

POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION RELATED TO LOSS OF POWER DIESEL GENERATOR START INSTRUMENTATION (TAC NOS. MD0936 AND MD0937)

Dear Mr. Koehl:

By letter to the U.S. Nuclear Regulatory Commission (NRC) dated March 23, 2006, Nuclear Management Company, LLC (NMC) submitted a proposed amendment to the technical specifications (TSs) for the Point Beach Nuclear Plant, Units 1 and 2, to revise TS 3.3.4, Loss of Power (LOP) Diesel Generator (DG) Start and Load Sequence Instrumentation.

The NRC staff is reviewing your submittal and has determined that additional information is required to complete its review. The specific information requested is addressed in the enclosure to this letter. During a discussion with Mr. J. Gadzala of your staff on November 15, 2006, it was agreed that NMC would provide a response by December 29, 2006, to this request for additional information.

The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRCs goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-2296.

Sincerely,

/RA/

Carl F. Lyon, Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301

Enclosure:

Request for Additional Information cc w/encl: See next page

Mr. Dennis Koehl November 27, 2006 Site Vice President Point Beach Nuclear Plant Nuclear Management Company, LLC 6610 Nuclear Road Two Rivers, WI 54241-9516

SUBJECT:

POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION RELATED TO LOSS OF POWER DIESEL GENERATOR START INSTRUMENTATION (TAC NOS. MD0936 AND MD0937)

Dear Mr. Koehl:

By letter to the U.S. Nuclear Regulatory Commission (NRC) dated March 23, 2006, Nuclear Management Company, LLC (NMC) submitted a proposed amendment to the technical specifications (TSs) for the Point Beach Nuclear Plant, Units 1 and 2, to revise TS 3.3.4, Loss of Power (LOP) Diesel Generator (DG) Start and Load Sequence Instrumentation.

The NRC staff is reviewing your submittal and has determined that additional information is required to complete its review. The specific information requested is addressed in the enclosure to this letter. During a discussion with Mr. J. Gadzala of your staff on November 15, 2006, it was agreed that NMC would provide a response by December 29, 2006, to this request for additional information.

The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRCs goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-2296.

Sincerely,

/RA/

Carl F. Lyon, Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301

Enclosure:

Request for Additional Information cc w/encl: See next page DISTRIBUTION:

PUBLIC LPL3-1 R/F RidsNrrDorlLple RidsNRRPMFLyon RidsNrrLATHarris RidsAcrsAcnwMailCenter RidsOgcRp RidsRgn3MailCenter RidsNrrDorlDpr RidsNrrDeEicb RidsNrrDeEeeb OChopra, NRR SMazumdar, NRR ADAMS Accession Number: ML063200086 OFFICE LPL3-1/PM LPL3-1/LA LPL3-1/BC NAME FLyon:ca THarris LRaghavan DATE 11/28/06 11/22/06 11/27/06 OFFICIAL RECORD COPY

Point Beach Nuclear Plant, Units 1 and 2 cc:

Jonathan Rogoff, Esquire Mr. Jeffery Kitsembel Vice President, Counsel & Secretary Electric Division Nuclear Management Company, LLC Public Service Commission of Wisconsin 700 First Street P.O. Box 7854 Hudson, WI 54016 Madison, WI 53707-7854 Mr. F. D. Kuester Nuclear Asset Manager President & Chief Executive Officer Wisconsin Electric Power Company WE Generation 231 West Michigan Street 231 West Michigan Street Milwaukee, WI 53201 Milwaukee, WI 53201 Michael B. Sellman Regulatory Affairs Manager President and Chief Executive Officer Point Beach Nuclear Plant Nuclear Management Company, LLC Nuclear Management Company, LLC 700 First Street 6610 Nuclear Road Hudson, MI 54016 Two Rivers, WI 54241 Douglas E. Cooper Mr. Ken Duveneck Senior Vice President & Chief Nuclear Town Chairman Officer Town of Two Creeks Nuclear Management Company, LLC 13017 State Highway 42 700 First Street Mishicot, WI 54228 Hudson, WI 54016 Chairman Site Director of Operations Public Service Commission Nuclear Management Company, LLC of Wisconsin 6610 Nuclear Road P.O. Box 7854 Two Rivers, WI 54241 Madison, WI 53707-7854 Regional Administrator, Region III U.S. Nuclear Regulatory Commission Suite 210 2443 Warrenville Road Lisle, IL 60532-4351 Resident Inspector's Office U.S. Nuclear Regulatory Commission 6612 Nuclear Road Two Rivers, WI 54241 November 2005

REQUEST FOR ADDITIONAL INFORMATION POINT BEACH NUCLEAR POWER PLANT, UNITS 1 AND 2 DOCKET NOS. 50-266 AND 50-301 In reviewing the Nuclear Management Company, LLCs (NMC) submittal dated March 23, 2006 (Agencywide Documents Access and Management System Accession No. ML060900452), the Nuclear Regulatory Commission (NRC) staff has determined that the following information is needed in order to complete its review:

The license amendment request (LAR) proposes the following Technical Specification (TS) change for the Point Beach Nuclear Plant, Units 1 and 2, TS Surveillance Requirement (SR) 3.3.4.3.b from:

Perform Channel Calibration for 4.16 kV degraded voltage Allowable Value > 3937 V with a time delay of < 6.47 seconds (with SI [safety injection] signal present) and < 54 seconds (without SI signal present) to:

Perform Channel Calibration for 4.16 kV degraded voltage Allowable Value > 3937 V with a time delay of < 5.68 seconds (bus degraded voltage relay) and < 39.14 seconds (bus time delay relay)

To support the NRC assessment of the acceptability of the LAR in regard to setpoint changes, provide the following for each setpoint to be added or modified:

1. Setpoint Calculation Methodology: Provide documentation (including sample calculations) of the methodology used for establishing the limiting setpoint (or nominal trip setpoint, NSP) and the limiting acceptable values for the as-found and as-left setpoints as measured in periodic surveillance testing as described below. Indicate the related analytical limits and other limiting design values (and the sources of these values) for each setpoint.

2. Safety Limit (SL)-related Determination: Provide a statement as to whether or not the setpoint is a limiting safety system setting (LSSS) for a variable on which a safety limit (SL) has been placed as discussed in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(c)(1)(ii)(A). Such setpoints are described as SL-Related in the discussions that follow. In accordance with 10 CFR 50.36(c)(1)(ii)(A), the following guidance is provided for identifying a list of functions to be included in the subset of LSSSs specified for variables on which SLs have been placed as defined in Standard Technical Specifications (STS) Sections 2.1.1, Reactor Core SLs and 2.1.2, Reactor Coolant System Pressure SLs. This subset includes automatic protective devices in TSs for specified variables on which SLs have been placed that: (1) initiate a reactor trip or (2) actuate safety systems. As such, these variables provide protection against violating reactor core SLs, or reactor coolant system pressure boundary SLs.

An example of instrument functions that might have LSSSs included in this subset in

accordance with the plant-specific licensing basis, is the pressurizer pressure reactor trip. For each setpoint, or related group of setpoints that NMC determines not to be SL-related, explain the basis for this determination.

3. For setpoints that are determined to be SL-related: The NRC letter to the nuclear energy Institute Setpoint Methods Task Force dated September 7, 2005 (ADAMS Accession No. ML052500004), describes setpoint-related TS (SRTS) that are acceptable to the NRC for instrument settings associated with SL-related setpoints.

Specifically: Part A of the enclosure to the letter provides limiting conditions for operation notes to be added to the TS, and Part B includes a check list of the information to be provided in the TS Bases related to the proposed TS changes.

a. Describe whether and how NMC plans to implement the SRTS suggested in the September 7, 2005 letter. If NMC does not plan to adopt the suggested SRTS, then explain how NMC will ensure compliance with 10 CFR 50.36, Technical specifications, by addressing Items 3b and 3c, below.

b. As-Found Setpoint evaluation: Describe how surveillance test results and associated TS limits are used to establish operability of the safety system.

Show that this evaluation is consistent with the assumptions and results of the setpoint calculation methodology. Discuss the plant corrective action processes (including plant procedures) for restoring channels to operable status when channels are determined to be inoperable or operable but degraded. If the criteria for determining operability of the instrument being tested are located in a document other than the TS (e.g., plant test procedure) explain how the requirements of 10 CFR 50.36 are met.

c. As-Left Setpoint control: Describe the controls employed to ensure that the instrument setpoint is, upon completion of surveillance testing, consistent with the assumptions of the associated analyses. If the controls are located in a document other than the TS (e.g., plant test procedure) explain how the requirements of 10 CFR 50.36 are met.

4. For setpoints that are not determined to be SL-related: Describe the measures to be taken to ensure that the associated instrument channel is capable of performing its specified safety functions in accordance with applicable design requirements and associated analyses. Include in the discussion information on the controls NMC employs to ensure that the as-left trip setting after completion of periodic surveillance is consistent with NMCs setpoint methodology. Also, discuss the plant corrective action processes (including plant procedures) for restoring channels to operable status when channels are determined to be inoperable or operable but degraded. If the controls are located in a document other than the TS (e.g., plant test procedure), describe how it is ensured that the controls will be implemented.

5. NMCs justifications for changing the wording in SR 3.3.4.3b from with SI signal present to bus degraded voltage relay and from without SI signal present to bus time delay relay, is not clear to the NRC staff. The NRC staff believes that the current wording is clear and is consistent with Branch Technical Position PSB-1 which clearly

requires two-time delay functions one with and the other without SI signal. Provide additional justification for this proposed change.

REFERENCES:

1. Letter from Patrick L. Hiland, NRC, to NEI Setpoint Methods Task Force, "Technical Specification for Addressing Issues Related to Setpoint Allowable Values," dated September 7, 2005, available on the NRC public website in the Agency Documents Access and Management System (ADAMS), Accession No. ML052500004.

2. Letter from Bruce A. Boger, NRC, to Alexander Marion, "Instrumentation, Systems, and Automatic Society (ISA) S67.04 Methods for Determining Trip Setpoints and Allowable Values for Safety-Related Instrumentation," dated August 23, 2005, ADAMS Accession No. ML051660447.

3. Letter from James A. Lyons, NRC, to Alex Marion, Nuclear Energy Institute (NEI),

"Instrumentation, Systems, and Automation Society S67.04 Methods for Determining Trip Setpoints and Allowable Values for Safety-Related Instrumentation," dated March 31, 2005, ADAMS Accession No. ML050870008.

4. NRC Regulatory Issue Summary 2006-17.