ML12335A232

Request for Additional Information, Round 2, Amendment Request to Adopt National Fire Protection Association Standard 805, Performance-Based Standard for Fire Protection for LWR Generating Plants (2001 Edition)

ML12335A232
Person / Time
Site:	Callaway
Issue date:	12/11/2012
From:	Lyon C Plant Licensing Branch IV
To:	Heflin A Union Electric Co
Lyon C
References
TAC ME7046
Download: ML12335A232 (15)

Text

UNITED STA"rES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 11, 2012 Mr. Adam C. Heflin Senior Vice President and Chief Nuclear Officer Union Electric Company P.O. Box 620 Fulton, MO 6S2S1

SUBJECT:

CALLAWAY PLANT, UNIT 1 - REQUEST FOR ADDITIONAL INFORMATION, ROUND 2, RE: ADOPTION OF NATIONAL FIRE PROTECTION ASSOCIATION STANDARD NFPA 80S (TAC NO. ME7046)

Dear Mr. Heflin:

By application dated August 29,2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML112420020), to the U.S. Nuclear Regulatory Commission (NRC), as supplemented by letters dated November 9, 2011 (ADAMS Accession No. ML113140044), April 17, 2012 (ADAMS Accession No. ML12108A239), and July 12, 2012 (ADAMS Accession No. ML12194A624), Union Electric Company (dba Ameren Missouri, the licensee) submitted a license amendment request to transition the fire protection licensing basis at the Callaway Plant, Unit 1, from Title 10 of the Code of Federal Regulations (10 CFR),

Section S0.48(b), [Appendix R], to 10 CFR S0.48(c), "National Fire Protection Association Standard NFPA 80S."

The NRC staff has determined that additional information, as requested in the enclosure, is needed to complete its review. Please provide a response to the questions within 60 days of the date of this letter. Review of your application is ongoing and additional questions may be forthcoming. If circumstances result in the need to revise the requested response date, please contact me at 301-41S-2296 or via e-mail at Fred.Lyon@nrc.gov.

Sincerely, Carl F. Lyon, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. SO-483

Enclosure:

As stated cc w/encl: Distribution via Listserv

REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST TO ADOPT NATIONAL FIRE PROTECTION ASSOCIATION STANDARD NFPA 805 UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT 1 DOCKET NO. 50-483 By application dated August 29, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML112420022), to the U.S. Nuclear Regulatory Commission (NRC), as supplemented by letters dated November 9, 2011 (ADAMS Accession No. ML113140044), April 17, 2012 (ADAMS Accession No. ML12108A239), and July 12, 2012 (ADAMS Accession No. ML12194A624), Union Electric Company (dba Ameren Missouri, the licensee) submitted a license amendment request to transition the fire protection licensing basis at the Callaway Plant from Title 10 of the Code of Federal Regulations (10 CFR), Section 50.48(b), [Appendix R], to 10 CFR 50.48(c), "National Fire Protection Association Standard NFPA 805." The NRC staff has determined that the additional information requested below is needed to complete its review.

Radioactive Release RAI 01 For liquid releases, Column 4 of Table E-2 indicates the Fire Areas where liquid effluents are collected in monitored tanks. For each area, Column 4 also states that Fire Pre-Plan steps provide reasonable assurance liquids are contained within the fire area. For each such area, please provide a qualitative assessment describing the following:

a. The type of fire most likely to occur in that fire area (e.g., electrical, transient combustibles, fuel);

b. The type and amount of radioactive contamination in the fire area;

c. The type offire suppression used in the area (e.g., water, foam);

d. The duration of anticipated fire fighting activities;

e. The anticipated amount of water to be generated;

f. The capability of the monitored tanks to contain the estimated amount of water to be generated;

g. Any specific actions/methods (e.g., temporary dikes, absorbent materials, directed fire hose spray) provided by the Fire Pre-Plan that are needed to ensure containment of the liquid effluents from this area.

Enclosure

-2 Radioactive Release RAI 02 For those areas where the heating ventilation and air conditioning (HVAC) systems are provided with "filtered monitored elevated release paths," Column 5 of Table E-2, states that Fire Pre Plan steps provide reasonable assurance gaseous products are contained within the building and fire area. For these areas, please provide the specific actions that will be used to contain the gaseous products.

Radioactive Release RAI 03 For Fire Areas A-21 and A-22, Column 5 of Table E-2 states that while HVAC systems in the Auxiliary Building (AB) are provided with filtered monitored elevated release paths that exhaust via the plant unit vent, these two areas are included within the main control room (MCR) HVAC boundary which is not monitored. The table also indicates Fire Pre-Plan steps provide reasonable assurance gaseous products are contained within the AB. Please provide clarification of the configuration of these areas and the specific steps that will be taken to provide reasonable assurance gaseous products are contained within the AB.

Radioactive Release RAI 04 For Fire Areas C-5 through C-8, Column 5 of Table E-2 indicates that these fire areas are not provided with filtered monitored elevated release paths. The table further indicates that if found contaminated, the smoke will be routed to the AB that is a filtered monitored elevated release path that exhausts via the plant vent and references Fire Pre-Plan steps that ensure that gaseous products are contained appropriately. Please provide the methods/criteria used to evaluate the presence of contamination in the smoke and specific steps in the Fire Pre-Plan that provides reasonable assurance of the containment of gaseous products within the building.

Radioactive Release RAI 05 For Areas TB-1 on page E-31 , and YD-1 on page E-32, Columns 4 (Liquid Effluents) and 5 (Gaseous Effluents) of Table E-2 reference Callaway Plant Calculation KC-43, "NFPA 805 Code Comparison." Section 4.4 for evaluations performed for specific locations within the Non Radiologically Controlled Areas (RCAs) of TB-1 and YD-1. For these two areas, a bounding analysis, Callaway Plant Calculation HPCI 10-04, is also referenced in Column 7 (Conclusions) of the table. Column 7 indicates that that the bounding analysis and administrative controls along with Fire Pre-Plan steps limits radiation release due to direct effects of fire suppression activities. It also states that in addition to compliance with NFPA 805 Sections 4.2.3 and 4.4.4, it satisfies the performance requirements of NFPA 805 for radiation release. Please provide copies of the referenced documents, administrative controls and Fire Pre-plan steps and provide justification as to how the release will not only not exceed the limits in NFPA 805 but also not exceed the instantaneous gaseous effluent dose rate limits and the liquid effluent concentration limits in the plant's Technical Specifications (TS) (Reference FAQ-09-0056, ADAMS Accession No. ML102920405.)

- 3 Probabilistic Risk Assessment RAI 24 Please describe how the evaluation includes the possible increase in heat release rate (HRR) caused by the spread of a fire from the ignition source to other combustibles. Please summarize how suppression is included in the evaluation.

Probabilistic Risk Assessment RAI 25 Transient fires should at a minimum be placed in locations within the plant physical access units (PAUs) where conditional core damage probabilities (CCDPs) are highest for that PAU (Le., at "pinch points"). Pinch points include locations of redundant trains or the vicinity of other potentially risk-relevant equipment, including the cabling associated with each. Transient fires should be placed at all appropriate locations in a PAU where they can threaten pinch points.

Hot work should be assumed to occur in locations where hot work is a possibility, even if improbable (but not impossible), keeping in mind the same philosophy. Please describe how transient and hot-work fires are distributed within the PAUs. In particular, please identify the criterion that determines where an ignition source is placed within the PAUs. Also, if there are areas within a PAU where no transient or hot-work fires are located since those areas are considered inaccessible, please define the criteria used to define "inaccessible." Note that an inaccessible area is not the same as a location where fire is simply unlikely, even if highly improbable.

Probabilistic Risk Assessment RAI 26 The transition report describes and justifies an initial coping time of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, after which, actions are necessary to maintain safe and stable beyond 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. Please provide a discussion of the actions necessary during and beyond 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> to maintain safe and stable conditions beyond 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> such as refilling fluid tanks or re-aligning systems.

a. Please describe whether the risk analysis models are run for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the risk analysis models are run for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, describe whether they include all required actions between 10 and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. Please evaluate quantitatively or qualitatively the risk associated with the failure of actions and equipment necessary to extend safe and stable beyond 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> (or 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, if at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is fully modeled in the probabilistic risk assessment (PRA>> given the post-fire scenarios during which they may be required.

Probabilistic Risk Assessment RAI 27 Please describe how fire-induced instrument failure (including no readings, off-scale readings, and incorrect/misleading readings) is addressed in the fire human reliability analysis (HRA) for new operator actions (not including post MCR abandonment which are addressed elsewhere) that have been credited in the risk estimates.

- 4 Probabilistic Risk Assessment RAI 28 Please identify if any variance from deterministic requirements (VFDRs) in the license amendment request (LAR) involved performance-based evaluations of wrapped or embedded cables. If applicable, please describe how wrapped or embedded cables were modeled in the fire PRA (FPRA) including assumptions and inSights on how the PRA modeling of these cables contributes to the VFDR delta-risk evaluations.

Probabilistic Risk Assessment RAI 29 Please identify any changes made to the internal events PRA (lEPRA) or FPRA since the last full-scope peer review of each of these PRA models that are consistent with the definition of a "PRA upgrade" in American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS)-RA-Sa-2009, "Standard for Level1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," as endorsed by NRC Regulatory Guide (RG) 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities." Also, please address the following:

a. If any changes are characterized as a PRA upgrade, please identify if a focused scope peer review was performed for these changes consistent with the guidance in ASME/ANS-RA-Sa-2009, as endorsed by RG 1.200, and describe any findings from that focused-scope peer review and the resolution of these findings.

b. If a focused-scope peer review has not been performed for changes characterized as a PRA upgrade, please describe what actions will be implemented to address this review deficiency.

Probabilistic Risk Assessment RAI 30 Please identify any plant modifications (implementation items) in Attachment S of the LAR that have not been completed but which have been credited directly or indirectly in the change-in risk estimates provided in Attachment W. When the effects of a plant modification have been included in the PRA before the modification has been completed, the models and values used in the PRA are necessarily estimates based on current plans. The as-built facility after the modification is completed may be different than the plans. Please add an implementation item that, upon completion of all PRA credited implementation items, verifies the validity of the reported change-in-risk. This item should include a plan of action should the as-built change-in risk exceed the estimates reported in the LAR Probabilistic Risk Assessment RAI 31 Please describe whether changes to the seismicities as a result of the United States Geological Survey (USGS) re-evaluation for the central and eastern United States (USGS, "2008 NSHM Gridded Data, Peak Ground Acceleration"), based on reanalysis of the New Madrid earthquakes, were considered in determining the applicability of the seismic-fire interaction analysis performed for the individual plant examination for external events (IPEEE) to the

- 5 current state of seismic-fire interactions. If not, please describe how the changes would affect the conclusions made regarding seismic-fire interactions.

Probabilistic Risk Assessment RAI 32 Please discuss the calculation of the frequencies of transient and hot-work fires. Please characterize the use of the influence factors for maintenance, occupancy, and storage, noting if the rating "3" is the most common, as it is intended to be representative of the "typical" weight for each influence factor. It is expected that the influence factor for each location bin associated with transient or hot-work fires will utilize a range of influence factors about the rating "3,"

including the maximum 10 (or 50 for maintenance) and, if appropriate, even the rating "0." Note that no PAU may have a combined weight of zero unless it is physically inaccessible, administrative controls notwithstanding. In assigning influence factor ratings, those factors for the Control/Auxiliary/Reactor Building are distinct from the Turbine Building; thus, the influence factor ratings for each location bin are to be viewed according to the bin itself.

Probabilistic Risk Assessment RAI 33 By letter dated July 12, 2012, the licensee responded to PRA RAI 01(c) regarding the effect of updated common cause failure (CCF) probabilities from the recent lEPRA update, and stated that, despite the presence of increases in some CCF probabilities, "a succinct sensitivity study for each CCF value to determine its individual effect was not possible within the time constraints of this RAI response ... The updated CCFs are scheduled to be incorporated as part of the next fire PRA revision." While this fact and observation (F&O) (1-13) is "only" a suggestion with respect to the lEPRA from the recent focused-scope peer review, it remains unknown as to the potential effect on the conclusions made for transition based on the FPRA. Please provide, at a minimum, a bounding analysis to estimate the potential effect upon fire core damage frequency (CDF), large early release frequency (LERF), delta (6) CDF and ~ LERF, as this is necessary to confirm that the conclusions remain the same.

Probabilistic Risk Assessment RAI 34 By letter dated April 17, 2012, the licensee responded to PRA RAI 02, and stated that The Fire PRA Peer Review is therefore consistent with the clarifications in RG 1.200, Rev. 2." If correct, please add the following "... the clarifications AND QUALIFICATIONS in ..." If not, please describe and justify why it is not consistent with "qualifications."

Probabilistic Risk Assessment RAI 35 Please describe how CDF and LERF are estimated in the MCR abandonment scenarios.

Please clarify if any fires outside of the MCR cause MCR abandonment because of loss of control and/or loss of control room habitability. Please clarify if "screening" values for post MCR abandonment are used (e.g., conditional core damage probability (CCDP) of failure to successfully switch control to the Primary Control Station and achieve safe shutdown of 0.1) or if detailed human error analyses been completed for this activity. Please justify any screening value used. The justification should provide the results of the human failure event (HFE) quantification process described in Section 5 of NUREG-1921, "EPRIINRC-RES Fire Human

-6 Reliability Analysis Guidelines, Final Report," July 2012 (ADAMS Accession No. ML 123326A104), including the following:

a. The results of the feasibility assessment of the operator action(s) associated with the HFEs, specifically addressing each of the criteria discussed in Section 4.3 of NUREG-1921.

b. The results of the process in Section 5.2.8 of NUREG-1921 for assigning scoping human error probabilities (HEPs) to actions associated with the use of alternate shutdown, specifically addressing the basis for the answers to each of the questions asked in the Figure 5-5 flowchart.

c. The results of a detailed HRA quantification, per Section 5.3 of NUREG-1921, if the screening CCDP is determined to not be bounding.

Fire Modeling RAI 03.01 NFPA 805, Section 2.7.3, "Quality," describes requirements for fire modeling calculations, such as acceptable models, limitations of use, validation of models, defining fire scenarios, etc. This description includes justification of model input parameters, as it is related to limitations of use and validation.

a. At the time of the site audit, additional fire modeling analysis and the reports describing the results of this analysis for Fire Area C-1 ("R1984-001-001 Fire Dynamic Simulator (FDS) Analysis to Support Detailed Fire Modeling," and "KC-57 Fire Modeling Report for Fire Area C-1") were not officially completed.

Since that time, these reports have been completed and reviewed. The latter resulted in the following questions:

i. In Section C1.4.2 of "R1984-001-001, FDS Analysis to Support Detailed Fire Modeling," there is a discussion of normalized parameters and their range of validity, per NUREG-1824, 'Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications," May 2007 (ADAMS Accession No. ML071650546), and draft NUREG-1934, "Nuclear Power Plant Fire Modeling Application Guide (NPP FIRE MAG)." The normalized parameter related to Radial Distance Relative to Fire Diameter is said not to be applicable since (1) 'The radiation heat flux to the HOPE [high density polyethylene] pipes is based on the temperature of the localized hot gas layer, not the fire; therefore, the heat flux from the fire within this model is not used to fail the HOPE pipes." and (2) " ... due to physical obstructions within the model, the radiant heat from the flame does not directly impact the HOPE pipes."

Please describe why this normalized parameter is not applicable.

ii. Section C1.5.7 of "R1984-001-001, FDS Analysis to Support Detailed Fire Modeling," states that a design fire of 69 kilowatts (kW) was used in the analysis. Section 7.3.2 of the document entitled, "KC-57 Fire

-7 Modeling Report for Fire Area C-1," provides justification for this value.

This justification is qualitative and it is not clear how the recommended 98th percentile value of 317 kW was quantitatively scaled to 69 kW.

Please describe whether this value was determined based on the method discussed in Section G.5 of NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, Final Report,"

September 2005 (ADAMS Accession Nos. ML052580075 and IVIL052580118, for Volumes 1 and 2, respectively). If so, please describe whether a representative distribution function of this point value of the HRR was developed or was this 'bounding' point value used in conjunction with a severity factor of 1.0 in the Fire PRA. Please provide clarification for the justification that a transient fire in excess of 69 kW cannot occur in this fire zone (Room 3101 in Fire Area C1).

iii. Section C1.5.7 of "R1984-001-001, FDS Analysis to Support Detailed Fire Modeling," states that 32 additional pounds of Class A or 13 additional pounds of Class B fire load would be necessary to transition from the maximum expected fire scenario (MEFS) HRR to the limiting fire scenario (LFS) HRR. Please provide additional clarification for how these estimates were made. Please explain what administrative controls are, or will be, in place to ensure these relatively small additional fire loads will not be present in this particular space. In addition, provide justification for the HRR of the IVIEFS used in the analysis.

iv. For the HRRs used in the analysis, a t2 profile was used for a transient fire that reached its peak at 8 minutes, as recommended in FAQ-08-0052. The time to peak HRR used in the analysis is recommended for common trash cans such as plastic or metal receptacles. The suggested time to peak HRR for trash fires that are not contained in plastic or receptacles is 2 minutes, which is more representative of a uno storage" area trash fire. Additionally, the suggested time to peak HRR for an oil spill (Class B) fire is 0 minutes.

Please explain what administrative controls are, or will be, put in place to prevent loose trash and liquid fuel spill fires in this area.

b. Independent fire modeling was conducted for various fire areas using the associated fire models. Based on independent use of the FDTs for fire scenario C-31.3618-8, the reported plume radius appears to have been calculated for thermoplastic damage criteria, as opposed to thermoset criteria. Please provide confirmation that this error does not have an adverse effect on the results and conclusions of the analysis.

c. The detailed fire modeling reports of several fire areas (e.g., A11, C1, C21, and C31) refer to the MEFS and the LFS. The terms MEFS and LFS are typically used when fire modeling is performed to support performance-based evaluations in accordance with NFPA 805, Section 4.2.4.1. However, Section 4.5.1.2 in the

- 8 LAR states that "Fire modeling was performed as part of the FPRA development (NFPA 805 Section 4.2.4.2)." Please provide confirmation that this statement in the LAR is correct and that no fire modeling was performed to support compliance with NFPA 80S. Section 4.2.4.1.

Please explain and justify: (1) the intended use and definition of the terms MEFS and LFS in this capacity, which appear to differ from how these terms are used and defined in NFPA 805 Section 4.2.4.1; and (2) how these terms were applied with regard to detailed fire modeling in support of the FPRA. For example. a 98th or 100th percentile fire scenario in accordance with NFPA 805 Section 4.2.4.2 might be used to serve a similar purpose to the MEFS in accordance with NFPA 805 Section 4.2.4.1.

Fire Protection Engineering RA115.01 In a letter dated April 17. 2012, the licensee responded to Fire Protection Engineering RAI 15 regarding high hose station pressures. and discussed the previous removal of pressure reducers as a result of Unresolved Item (URI) #483/87018-03. Based on this response. the referenced URI. and Calculation KC-27. additional information is required:

a. Typically. a fire hose is service tested to 150 pounds per square inch (psi). The letter referenced above indicates operating pressures up to 160 psi. Please clarify that the service testing pressure for fire hoses adequately accounts for this higher pressure (e.g., service testing exceeds the maximum operating pressure usually 110 percent or 50 psi over the maximum operating pressure).

b. The closure of URI #483/87018-03 required the posting of signs on hose stations with high pressures. Please clarify the continued adequacy of such signs for operating pressures up to 160 psi.

c. Please clarify the following discrepancy: Calculation KC-27 refers to URI #483/870113-03; however. the above-referenced letter refers to URI #483/87018-03.

Fire Protection Engineering RAI16 LAR Table 1-1: Power Block Definition lists power block structures within the Owner Controlled Area that were determined to contain equipment required to meet the nuclear safety and radioactive release performance criteria. In comparing those structures listed in Frequently Asked Question (FAQ) 06-0019 (ADAMS Accession No. ML080510224). and NEI 04-02, "Guidance for Implementing a Risk-Informed. Performance Based Fire Protection program Under 10 CFR50.48(c)," Section K. the NRC staff noticed certain structures were not listed in LAR Table 1-1.

Please ensure those structures listed within NEI 04-02 Attachment K and FAQ-06-0019 are accounted for as either within or not within the power block (e.g., Service Building, Water Treatment, Intake Structure, etc.). In addition, please clarify that the Hot Machine Shop, if one

- 9 exists, and the YARD should also be within the power block. For the YARD, please identify the specific structures or equipment to be within the power block, which may include fire pump house, transformers, YARD equipment managed by plant staff, warehouses, and other significant structures containing equipment required for operations. Please revise LAR Table 1-1 as necessary.

Fire Protection Engineering RAI17 NFPA 805 Section 3.3.5.1 requires cables located above suspended ceilings to be listed for plenum use, routed in armored cable, routed in metallic conduit, or routed in cable trays. LAR Attachment L, Approval Request 2, has identified certain areas and cables that do not meet NFPA 805 Section 3.3.5.1 and provided justifications to deviate from such requirements.

a. Please clarify that the cables that do not meet the NFPA 805 Section 3.3.5.1 criteria are not in the vicinity of nuclear safety capability systems and equipment.

b. Please clarify if there are any fire detection systems installed above suspended ceilings in these areas of concern.

For those cables that do not meet the NFPA 805 Section 3.3.5.1 criteria, please describe the separation between the power, control, and lighting cables and the telephone/data/communication cables, if any.

c. The Basis for Request section states that "for the cables that do meet the NFPA 805 Section 3.3.5.1 criteria, the majority meet one of the cable qualifications listed within FAQ 06-0022, Rev. 3." Please clarify if the statement was meant to say, "for those cables that do not meet the NFPA 805 Section 3.3.5.1 criteria ... " If not, then clarify the meaning and significance of this sentence.

d. The Basis for Request section states that "plant procedures will be revised to ensure future exposed cables installed above the suspended ceilings meet one of the cable qualifications found acceptable in FAQ 06-0022 Rev. 3." Please clarify why future installations are not required to meet NFPA 805 Section 3.3.5.1 requirements.

e. The regulations in 10 CFR 50.48(c)(2)(vii) require that radioactive release performance measures must be addressed by the request as well as nuclear safety performance measures. Please revise the request to address the radioactive release performance measures (goals, objectives, and performance criteria). Please include in your discussion how the non-fire water use could potentially impact firefighting efforts and how this might impact liquid and gaseous releases.

f. The Safety Margin section states that "the safety margin inherent in the analysis for the fire event has been preserved." Please clarify that this performance

- 10 based method does not change the assumptions and limitations of the analytical methods used in the development of the FPP.

Programmatic RAI 01 Section 2.7.3, "Quality," of NFPA 805 has the following specific requirements:

2.7.3.1: Each analysis, calculation, or evaluation performed shall be independently reviewed.

2.7.3.2: Each calculational model or numerical method used shall be verified and validated through comparison to test results or comparison to other acceptable models.

2.7.3.3: Acceptable engineering methods and numerical models shall only be used for applications to the extent these methods have been subject to verification and validation. These engineering methods shall only be applied within the scope, limitations, and assumptions prescribed for that method.

2.7.3.4: Cognizant personnel who use and apply engineering analysis and numerical models (e.g., fire modeling techniques) shall be competent in that field and experienced in the application of these methods as they relate to nuclear power plants, nuclear power plant fire protection, and power plant operations.

2.7.3.5: An uncertainty analysis shall be performed to provide reasonable assurance that the performance criteria have been met."

Section 4.7.3 of the Transition Report states that the requirements of NFPA 805 Sections 2.7.3.1 through 2.7.3.5 were followed for analyses supporting the transition to NFPA 805. However, no specific commitment has been made to comply with these requirements for future analyses. Please provide this commitment or define any alternative requirements that will be used for future analyses.

Safe Shutdown Analysis RAI 07 Databases and software that integrate fire protection program structure, system, and component data; fire modeling results, and PRA analyses (e.g., EPM-SAFE-PB and ARC) have a range of uses applicable to NFPA 805 implementation. These uses are subject to several NFPA 805 requirements including those that address determination of success paths; completion of the Nuclear Safety Capability Assessment (NSCA); the quality, configuration control, documentation, and verification and validation of analyses; and limitations of use. In addition, these databases and software can be used to facilitate integration of several aspects of NFPA 805 compliance. Specific applicable NFPA 805 requirements include:

NFPA 805 Section 2.2.9 "Plant Change Evaluation" states that: "In the event of a change to a previously approved fire protection program element, a risk informed plant change evaluation shall be performed and the results used as described in

- 11 2.4.4 to ensure that the public risk associated with fire-induced nuclear fuel damage accidents is low and that adequate defense-in-depth and safety margins are maintained."

NFPA 805, Section 2.2.11 "Documentation and Design Configuration Control" requires that: "The fire protection program documentation shall be developed and maintained in such a manner that facility deSign and procedural changes that could affect the fire protection engineering analysis assumptions can be identified and analyzed."

NFPA 805 Section 2.4.1 "Fire Modeling Calculations" requires: (2.4.1.1) "The fire modeling process shall be permitted to be used to examine the impact of the different fire scenarios against the performance criteria under consideration. n (2.4.1.2.1) "Only fire models that are acceptable to the authority having jurisdiction shall be used in fire modeling calculations." (2.4.1.2.2) "Fire models shall only be applied within the limitations of that fire mode!." (2.4.1.2.3) "The fire models shall be verified and validated."

NFPA 805 Section 2.4.3.3 regarding fire risk evaluations states: The PSA approach, methods, and data shall be acceptable to the AHJ. They shall be appropriate for the nature and scope of the change being evaluated, be based on the as-built and as-operated and maintained plant, and reflect the operating experience at the plant."

NFPA 805 Section 2.4.4, "Plant Change Evaluation" states: "A plant change evaluation shall be performed to ensure that a change to a previously approved fire protection program element is acceptable. The evaluation process shall consist of an integrated assessment of the acceptability of risk, defense-in-depth, and safety margins. The impact of the proposed change shall be monitored."

NFPA 805 Content requirements include:

(2.7.1.1) The analyses performed to demonstrate compliance with this standard shall be documented for each nuclear power plant (NPP). The intent of the documentation is that the assumptions be clearly defined and that the results be easily understood, that results be clearly and consistently described, and that sufficient detail be provided to allow future review of the entire analyses.

Documentation shall be maintained for the life of the plant and be organized carefully so that it can be checked for adequacy and accuracy either by an independent reviewer or by the AHJ."

(2.7.1.2) "A fire protection program design basis document shall be established based on those documents, analyses, engineering evaluations, calculations, and so forth that define the fire protection deSign basis for the plant. As a minimum, this document shall include fire hazards identification and nuclear safety capability assessment, on a fire area basis, for all fire areas that could affect the nuclear safety or radioactive release performance criteria defined in Chapter 1."

- 12 (2.7.1.3) "Detailed information used to develop and support the principal document shall be referenced as separate documents if not included in the principal document."

NFPA 805 configuration control requirements include:

(2.7.2.1) "The design basis document shall be maintained up-to-date as a controlled document. Changes affecting the design, operation, or maintenance of the plant shall be reviewed to determine if these changes impact the fire protection program documentation."

(2.7.2.2) "Detailed supporting information shall be retrievable records. Records shall be revised as needed to maintain the principal documentation up-to-date."

Finally, NFPA 805 quality requirements apply to use of integration databases and software:

(2.7.3.1) "Each analysis, calculation, or evaluation performed shall be independently reviewed."

(2.7.3.2) "Each calculational model or numerical method used shall be verified and validated through comparison to test results or comparison to other acceptable models."

(2.7.3.3) "Acceptable engineering methods and numerical models shall only be used for applications to the extent these methods have been subject to verification and validation. These engineering methods shall only be applied within the scope, limitations, and assumptions prescribed for that method."

(2.7.3.4) "Cognizant personnel who use and apply engineering analysis and numerical models (e.g., fire modeling techniques) shall be competent in that field and experienced in the application of these methods as they relate to nuclear power plants, nuclear power plant fire protection, and power plant operations."

"An uncertainty analysis shall be performed to provide reasonable assurance that the performance criteria have been met."

The NRC staff notes that, given the broad range of requirements applicable to use of integration databases and software, the Transition Report provided insufficient details for the staff to complete its review of the various areas affected by this software, Therefore, the staff requests that the following additional information be provided, a, A description of how the post transition change evaluation process will ensure that the potential interfaces between integration databases and software and other databases and analyses (e.g" the cable and raceway database, the NSCA, the fire PRA, and fire modeling) are evaluated and updated, as appropriate,

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b. A description of the process that will be employed to ensure that integration databases and software are maintained in accordance with documentation and design configuration control processes and procedures.

c. A description of the process and procedures that will be used to ensure that integration database and software analyses are conducted and/or updated by persons properly trained and experienced in its use.

d. A description of the processes and procedures that will be used to ensure that integration database and software analyses comply with NFPA 805 fire modeling, content, and quality control requirements.

December 11,2012 Mr. Adam C. Heflin Senior Vice President and Chief Nuclear Officer Union Electric Company P.O. Box 620 Fulton, MO 6S2S1

SUBJECT:

CALLAWAY PLANT, UNIT 1 - REQUEST FOR ADDITIONAL INFORMATION, ROUND 2, RE: ADOPTION OF NATIONAL FIRE PROTECTION ASSOCIATION STANDARD NFPA 80S (TAC NO. ME7046)

Dear Mr. Heflin:

By application dated August 29, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML112420020), to the U.S. Nuclear Regulatory Commission (NRC), as supplemented by letters dated November 9,2011 (ADAMS Accession No. ML113140044), April 17, 2012 (ADAMS Accession No. ML12108A239), and July 12,2012 (ADAMS Accession No. ML12194A624), Union Electric Company (dba Ameren Missouri, the licensee) submitted a license amendment request to transition the fire protection licensing basis at the Callaway Plant, Unit 1, from Title 10 of the Code of Federal Regulations (10 CFR),

Section S0.48(b), [Appendix R], to 10 CFR S0.48(c), "National Fire Protection Association Standard NFPA 80S."

The NRC staff has determined that additional information, as requested in the enclosure, is needed to complete its review. Please provide a response to the questions within 60 days of the date of this letter. Review of your application is ongoing and additional questions may be forthcoming. If circumstances result in the need to revise the requested response date, please contact me at 301-41S-2296 or via e-mail at Fred.Lyon@nrc.gov.

Sincerely, IRA by Lauren Kate Gibson forI Carl F. Lyon, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. S0-483

Enclosure:

As stated cc w/encl: Distribution via Listserv DISTRIBUTION:

PUBLIC RidsNrrLAJBurkhardt Resource RidsNrrDraAhpb Resource LPLIV Reading RidsNrrPMCallaway Resource RidsNrrDraApla Resource RidsAcrsAcnw_MailCTR Resource RidsOgcRp Resource RidsNrrDraAfpb Resource RidsNrrDorlLpl4 Resource RidsRgn4MailCenter Resource JRobinson. NRR/ORAlAFPB ADAMS Accession No.: ML12335A232 *email dated OFFICE NRRlDORULPL4/PM NRRlDORULPL4/LA AFPB/PM NRRlDORULP RlDORULPL4IPM MMarkley FLyon ( LKGibson NAME FLyon JBurkhardt (NKalyanam for) for)

DATE 12/7/12 12/6/12 17/12 12/11/12