ULNRC-05823, Clarification of Information Contained in Request for License Amendment to Adopt NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition) (LDCN 11-0012)

Clarification of Information Contained in Request for License Amendment to Adopt NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition) (LDCN 11-0012)

ML113140044
Person / Time
Site:	Callaway
Issue date:	11/09/2011
From:	Maglio S Ameren Missouri, Union Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
ULNRC-05823
Download: ML113140044 (8)

Text

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WAmeren Callaway Plant MISSOURI November 9, 2011 ULNRC-05823 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 10 CFR 50.90 Ladies and Gentlemen:

DOCKET NUMBER 50-483 CALLAWAY PLANT UNIT 1 UNION ELECTRIC CO.

FACILITY OPERATING LICENSE NPF-30 CLARIFICATION OF INFORMATION CONTAINED IN REQUEST FOR LICENSE AMENDMENT TO ADOPT NFPA 805, "PERFORMANCE-BASED STANDARD FOR FIRE PROTECTION FOR LIGHT WATER REACTOR GENERATING PLANTS (2001 EDITION>" (LDCN 11-0012)

Reference:

1) Ameren Missouri Letter ULNRC-05781, "Request for License Amendment to Adopt NFPA 805, 'Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)' (License Amendment Request LDCN 11-0012)," dated August 29, 2011 In the referenced letter, Union Electric Company ( dba Ameren Missouri) requested an amendment to Facility Operating License Number NPF-30 for the Callaway Plant pursuant to 10 CFR 50.90. The letter and its attachments were submitted for the purpose of adopting a new fire protection licensing basis based on NFPA 805 and which complies with the requirements in 10 CFR 50.48(a) and (c) and the guidance in Revision 1 ofRegulatory Guide (RG) 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants." Ameren Missouri's request is now under review by the NRC staff.

In a teleconference conducted on October 27, 2011, the NRC Staff requested that Ameren Missouri clarify certain information contained in the amendment application. The attachment to this letter accordingly contains the requested clarifications.

The provided clarifications do not impact the 10 CFR 50.92 evaluation of"No Significant Hazards Consideration" previously provided in the referenced application. Further, this letter makes no new commitments or changes to any commitments.

PO Box 620 Fulton, MO 65251 AmerenMissouri.com

ULNRC-05823 November 9, 2011 Page2 Ifthere are any questions, please contact Scott Maglio at 573-676-8719 or Roger Wink at 314-225-1561.

I declare under penalty of perjury that the foregoing is true and correct.

Sincerely, Executed on: - L l I C1 I toll

~:tt~~g~

Regulatory Affairs Manager RCW/nls

Attachment:

Request for Clarifying Information

ULNRC-05823 November 9, 20II Page3 cc: Mr. Elmo E. Collins, Jr.

Regional Administrator U.S. Nuclear Regulatory Commission Region IV 6I2 E. Lamar Blvd., Suite 400 Arlington, TX 76011-4I25 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 820 I NRC Road Steedman, MO 65077 Mr. Mohan C. Thadani (2 copies)

Senior Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 0-8B I Washington, DC 20555-2738

ULNRC-05823 November 9, 2011 Page4 Index and send hardcopy to QA File A160.0761 Hardcopy:

Certrec Corporation 4200 South Hulen, Suite 422 Fort Worth, TX 76109 (Certrec receives ALL attachments as long as they are non-safeguards and may be publicly disclosed.)

Electronic distribution for the following can be made via Responses and Reports ULNRC Distribution:

A. C. Heflin F. M. Diya C. 0. Reasoner ill L. H. Graessle S.M. Maglio T. B. Elwood R. C. Wink R. Holmes-Bobo NSFU3 Secretary Mr. Paul Parmenter, Director (SEMA)

Mr. Thomas Mohr, Senior REP Planner (SEMA)

Mr. John Campbell, REP Planner (SEMA)

Ms. Diane M. Hooper (WCNOC)

Mr. Tim Hope (Luminant Power)

Mr. Ron Barnes (APS)

Mr. Tom Baldwin (PG&E)

Mr. Wayne Harrison (STPNOC)

Ms. Linda Conklin (SCE)

Mr. John O'Neill (Pillsbury Winthrop Shaw Pittman LLP)

Mr. Dru Buntin (DNR)

Attachment to ULNRC-05823 Page I of4 Request for Clarifying Information

References:

1) Ameren Missouri Letter ULNRC-05781, "Request for License Amendment to Adopt NFPA 805, 'Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)' (License Amendment Request LDCN 11-0012)," dated August 29,2011

2) NRC Regulatory Guide 1.174 (Revision 1), "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," dated November 2002 (ADAMS Accession Number ML023240437)

During a teleconference held on October 27, 2011 between NRC staff and Callaway personnel, the Staff requested that information contained in the Transition Report of the above referenced license amendment request (LAR) be clarified to assist the Staff in completing the Callaway Plant NFP A 805 LAR Acceptance Review. Each clarification/information request is identified and addressed below.

1. In attachment U ofEnclosure 1 to ULNRC-05781, Callaway Plant stated that the internal events PRA underwent a gap assessment against the Capability Category II requirements of ASME RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," with ASME RA-Sa-2003 and ASME RA-Sb-2005 Addenda, ASME, 2005. The Staff requests additional information on the scope of the gap assessment (e.g. a review of resolutions for previous findings, a focused scope review or a full scope review) as well as clarification on which version of the ASME PRA Standard was used to evaluate the PRA during the gap assessment.

Response

Callaway Plant considers the Internal Events gap assessment conducted in 2006 and referenced in Attachment U of the LAR to meet the criteria of a full scope peer review against the requirements of ASME RA-Sb-2005.

The formal report for the subject gap assessment was titled "Callaway PRA Gap Analysis Report."

The assessment was titled in this way due to the generic use of the term 'gap assessment' (as commonly used at the time) to mean an assessment of performance relative to a standard. It was not intended to reflect the formal definition of a PRA gap assessment as the term is currently used.

The assessment that was performed was an independent, full-scope peer review of the entire PRA model using the most current available versions of applicable Standards. The following excerpts from the final report provide additional detail on the scope of the review as well as the standards applied:

" ... AmerenUE [now Ameren Missouri] has chosen to perform a gap analysis to identify the areas of the PRA which need to be strengthened in order to assure that the Callaway PRA conforms to all the existing standards in sufficient depth to address all currently envisioned applications."

Attachment to ULNRC-05823 Page 2 of4 "This report documents the results of the gap analysis conducted of the complete Callaway PRA model, data and documentation in accordance with the Category II requirements of the ASME Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, updated to include Addenda B [ASME RA-Sb-2005], the Category II requirements of ANSI/ANS-58.21-2003, "American National Standard External-Events PRA Methodology," and the expected requirements of the ANS Low Power and Shutdown PRA Standard (draft being written). The review was conducted by a team of three senior experts with experience in performing NEI PRA Certifications and pre-Certification reviews with support from a sister plant PRA staff member and experts in the areas of shutdown risk analyses and human reliability analyses."

The subject matter experts discussed above were contractors and an industry peer brought in to conduct the review, all of whom would have met the current peer review team qualification requirements of section 1-6.2 of the combined PRA standard (ASME/ANS RA-Sa-2009).

2. In attachment W ofEnclosure 1 to ULNRC-5781, Callaway Plant stated that total plant risk is not higher than 1E-4 for core damage frequency (CDF) or 1E-5 for large early release frequency (LERF). These values correspond to the baseline risk criteria specified in Regulatory Guide 1.174 (Reference 2) for small changes in risk resulting from plant specific changes to the licensing basis.

The Staff requests additional detail on this statement in the form of total CDF and LERF values, along with the values for individual significant contributors (e.g. internal events, fire, other external events) to these totals.

Response

Callaway Plant does not have quantified models for seismic or other non-fire external events.

Following the guidance provided by the NRC in NUREG-1407, Callaway Plant performed and submitted with the IPEEE a focused-scope seismic margins assessment (SMA) in accordance with EPRI NP-6041-SL. Callaway Plant has not developed a quantifiable seismic risk model since completion of the original SMA. However, in response to Generic Issue 199, "Generic Issue 199 (GI-199) Implications ofUpdated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants," the NRC developed a Safety/Risk Assessment ofGI-199 1 which quantified seismic risk using Callaway site-specific SMA results and updated seismic hazard curves. A Seismic CDF (SCDF) estimate was developed by integrating the mean seismic hazard curve and the mean plant-level fragility curve. This method, developed by Kennedy (1997), is discussed in Section 10.B.9 of ASME/ANS RA-Sa-2009 and has previously been used by the Staff in the resolution ofGI-194, "Implications ofUpdated Probabilistic Seismic Hazard Estimates," and is the basis for the seismic performance-based approach for determining the site Safe Shutdown Earthquake as described in Regulatory Guide 1.208.

1 Letter from Patrick Hiland to Brian Sheron dated September 2, 2010, "Generic Issue 199 (GI-199) Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants Safety/Risk Assessment"

Attachment to ULNRC-05823 Page 3 of4 SCDF estimates for Callaway Plant in this analysis were produced using two sets of mean seismic hazard curves generated by the following: (1) NRC, based on U.S. Geological Survey, 2008, and (2) Lawrence Livermore National Laboratory, 1994. The results are provided in Table 1, "Seismic Risk Estimates."

Table 1: Seismic Risk Estimates IPEEE Weakest Controlling Simple PGA 10Hz 5Hz 1Hz Max Weighted Link Curve Avg Avg Model Seismic Core-Dama~e Frequencies Usin~ 2008 USGS Seismic Hazard Curves 1.8E-6 1.9E-6 6.3E-7 1.5E-7 1.9E-6 10Hz l.lE-6 l.OE-6 2.0E-6 Seismic Core-Damage Frequencies Using 1994 LLNL Seismic Hazard Curves 2.3E-6 1.7E-6 1.8E-6 3.6E-7 2.3E-6 PGA 1.5E-6 1.4E-6 2.3E-6 Based on the results of this analysis as well as the results of the original seismic margins assessment, the contribution from seismic risk is minimal. The highest risk estimate from this analysis in conjunction with the internal events and fire risk estimates exhibit significant margin to the total risk thresholds specified in Regulatory Guide 1.174, as shown in Table 2, "Total Risk and Contributors." (See next page.)

Risk-informed technical assessments may generally be conducted using CDF, LERF, public dose (person-rem), or a combination of these risk metrics. The selection of the appropriate risk metric(s) to assess an issue depends on the specific nature of the issue being assessed. Seismic risk assessments involve the implications of probabilistic seismic hazard estimates that describe the distribution (frequency and size) of seismically induced site vibratory ground motions. Although each of the three risk metrics (CDF, LERF, and public dose) depends on the seismic hazard, SCDF is expected to be the most sensitive to changes in the seismic hazard. Thus, the seismic Safety/Risk Assessment did not generate seismic LERF estimates. Seismic LERF risk is expected to be essentially proportional to the SCDF which would result in only a small contribution to LERF. Given the margin between the LERF estimates and the risk thresholds, LERF contributions from seismic risk are considered negligible.

The IPEEE review concluded that the Callaway Plant design conforms to the screening criteria (1975 Standard Review Plan) and that no potential vulnerabilities from high winds and other external events exist that have not been included in the original design analysis. Therefore, these hazards were screened and the quantitative risk is considered to be negligible.

Attachment to ULNRC-05823 Page 4 of4 Table 2: Total Risk and Contributors CDF Internal Events 2.61E-5 /yr Fire 2.03E-5 /yr Seismic2 2.30E-6 /yr Other External Events Negligible Total CDF 4.87E-5 /yr LERF Internal Events 4.20E-7 /yr Fire 3.99E-7 /yr Seismic and Other External Events negligible Total LERF 8.19E-7 /yr

3. It appears that based on the description provided in section 4.6 of the Transition Report (Enclosure 1 of the LAR), the Callaway Plant Monitoring Program may be separate and independent from the Maintenance Rule program. If this is so, more detailed information is needed to explain how monitoring will be accomplished for Nuclear Safety systems, structures and components.

Response

Callaway Plant's NFPA 805 Monitoring Program will not be separate and independent from the Callaway Plant Maintenance Rule Program and will be developed in accordance with the final approved revision ofNRC FAQ 10-0059.

2 Seismic Core-Damage Frequencies Using 1994 LLNL Seismic Hazard Curves- Weakest Link Model