ML13261A114

R. E. Ginna Nuclear Power Plant, E-mail Re. NFPA 805 License Amendment Request, Audit RAIs

ML13261A114
Person / Time
Site:	Ginna
Issue date:	09/12/2013
From:	Thadani M Division of Operating Reactor Licensing
To:	Harding T Constellation Energy Nuclear Group
Thadani M
References
TAC MF1393
Download: ML13261A114 (27)

Text

From: Thadani, Mohan Sent: Thursday, September 12, 2013 8:50 AM To: Harding Jr, Thomas (Thomas.HardingJr@cengllc.com)

Cc: Robinson, Jay

Subject:

R.E. Ginna Nuclear Power Plant, NFPA 805 License Amendment Request, Audit RAIs, TAC No. MF1393 Attachments: Ginna Audit RAls.docx Tom:

By letter dated March 28, 2013, (ADAMS Accession No. ML13093A064), R.E. Ginna Nuclear Power Plant, LLC, (Ginna) submitted a license amendment request (LAR) (ADAMS Accession No. ML13093A065) to transition the fire protection licensing basis at Ginna, from Title 10 of the Code of Federal Regulations (CFR),

Section 50.48(b), to 10CFR50.48{c), National Fire Protection Association Standard NFPA 805 (NFPA 805).

The NRC staff has reviewed the information provided by Ginna and also participated in an audit from August 26 to August 29, 2013 and concluded that additional information is needed to complete the review.

Attached please find an advance copy of request for additional information (RAI) arising from the RE Ginna Nuclear Power Plant, NFPA 805 audit by the NRC staff, that was completed during August 26 - 29, 2013. Please review the RAI, and let me know by a phone call to schedule a clarification phone call within two weeks after the receipt of this email. This process was discussed with the licensee's staff during the audit. There will probably be some more general editing between now and the issuance of the final RAt. No additional RAI is planned at this time. The current schedule shows that we will send the final version of the RAI to the licensee by the end of September. We request a response to that request within 60 calendar days from the date of the final letter, or a justification from the licensee why the requested 60 days response date cannot be met.

Best regards, Mohan C Thadani Senior Project Manager FitzPatrick, Ginna, and Constellation Fleet Plant Licensing Branch 1-1 Division of Operating Reactor licensing Office of Nuclear Reactor Regulation (301) 415-1476 Mohan.Thadani@nrc.gov

~U.S.NRC v...... _ _ a.pI-.C'_

~,.",,QJ.w;;:;.;.;;;

1

REQUEST FOR ADDITIONAL INFORMATION VOLUNTARY FIRE PROTECTION RISK INITIATIVE R.E. GINNA NUCLEAR POWER PLANT, LLC R.E. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244 Fire Protection Engineering (FPE) Request for Additional The compliance statement for License Amendment Access and Management System (ADAMS) Accession Nu Table B-1, Section 3.3.6 is "complies with use of Association Standard 805, "Performance-Based Sta Reactor Electric Generating Plants" (NFPA 805),

Class A as determined by tests described in Tests of Roof Coverings". The compliance basi FM class I. .. ". The basis further describes the Class A rating versus the scope of for a FM Approval Standard 4470, "Ap Sheet, Built-Up Roof (BUR) and Liqu Noncombustible Roof Deck C materials may achieve an FM Class I of flame, intermittent flame ing how the requirements met FPE RAI 02 The compliance 1, Section 3.3.7 states that flammable gas the total content is less than 400 standard cubic tates that Standard 55, 2010 edition, "Compressed

" Section 10.1.1, indicates that the chapter "does not apply to ng a total hydrogen content of less than 400 scf, if each than Sft." Describe the configuration of flammable administrative controls used to ensure the volume of ow 400 scf.

FPE RAI 03 The compliance for LAR Attachment A, Table B-1, Section 3.4.1(c) is "complies".

NFPA 805, Section 3.4.1{c) speCifically requires the fire brigade leader and two brigade members to have sufficient training and knowledge of nuclear safety systems to understand the effects of fire and fire suppressants on nuclear safety performance criteria. The compliance basis for this element states that that the Brigade Captain and Backup Brigade Captain are Auxiliary Operators, but does not specify the details of the training and knowledge of these members.

Describe how the requirements of NFPA 805 Section 3.4.1{c) are met with regard to training and knowledge of the brigade leader and at least two of the brigade members.

An approach acceptable to the staff for meeting this training and knowledge requirement is provided in Regulatory Guide (RG) 1.189, "Fire Protection for Nuclear Power Plants," Rev. 2, Section 1.6.4.1, Qualifications:

"The brigade leader and at least two brigade members should have sufficient training in or knowledge of plant systems to understand the effects of fire and fire suppressants on safe-shutdown capability. The brigade leader should be competent to assess the potential safety consequences of a fire and advise control room personnel. Such competence by the brigade leader may be evidenced by . of an operator's license or equivalent knowledge of plant systems."

FPE RAI 04 The compliance basis for LAR Attachment A, Table B-1 is installed in Battery Room 'I B. LAR Attachment S, to evaluate the existing configuration of the wrap cables for 45 minutes and to modify the wrap Section 3.11.5 states that "electrical raceway 4 shall be capable of resisting the fire effects of tested in accordance with and shall the '.:1 f'r'o nir'.:1I Supplement 1." Provide additional rding the requirements of NFPA 805 Section 3. in the ERFBS duration is based on fire testing FPE RAt 05 LAR Attachment L, of the use of two portable gas fire heaters in the " ..o,:.n", from freezing during cold weather.

The basis for the ment credited for the Nuclear Safety Capability Assessm the motor fire pump and service water pumps, a the portable gas heaters, which places them 0 (ZOI) of the Describe the administrative controls that located at a minimum of 40 feet from the NSCA equipment, describe how the administrative controls of n installed, including any changes to bulk storage or Request 2, is for approval of wiring above suspended ceilings that may not comply with irements of NFPA 805, Section 3.3.5.1. The basis for the request indicates that there are multiple areas where suspended ceilings are located, however only the Control Room (CR) was discussed in detail. Provide a description of the other fire areas that contain wiring above suspended ceilings, including proximity to fire areas containing nuclear safety capability systems and equipment. Also, include a discussion on the type, use, and amount of wiring, proximity to combustibles, presence of ignition sources, as well as any fire detection or suppression features that may be installed.

FPE RAI 07 LAR Attachment L, Approval Request 3, is for approval of the use of wiring that may not comply with the requirements of NFPA 805, Section 3.3.5.3. The Approval Request states that the performance based (PB) method is specifically associated with the installation of video/communication/data cables. However, the basis for the request includes a discussion of power cables used for modifications such as the spent fuel pool bridge crane which was modified using cables others than those meeting the Institute of Electrical and Electronics Engineers (IEEE) Standard 383, "IEEE Standard for Qualifying Class 1E Electric Cables and Field Splices for Nuclear Power Generating Stations." It is unclear this request is for all types of cable or only video/communication/data cables. Clarify of the request. In addition, provide further justification for the acceptability of usi hat does not comply with NFPA 805, Section 3.3.5.3. Include a qualitative or evaluation, as well as a more detailed discussion of how safety margin (SM) and (DID) are maintained using this PB approach.

FPE RAI 08 LAR Attachment L, Approval Request 4, is for oil mist that results from pump/motor operation.

8. Further characterization tion of deposition, and the fire hazard associated with ng proximity to equipment necessary to meet nuclear clude the basis for acceptability.

b. What actions, if equipment surfaces (e.g., during

Safe Shutdown Analysis RAJ 01 NFPA 805, Section 1.3.1 states the nuclear safety goal that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition. LAR section 4.2.1.2 states that safe and stable conditions can be maintained indefinitely until a decision is made to transition to residual heat removal (RHR) cooling. The LAR summarizes the means to maintain safe and stable conditions for extended periods of time, including inventory control, decay heat removal, electrical systems, and diesel fuel supplies. Provide additional discussion of the actions necessary beyond 24-hours to meet the specific nuclear safety performance criteria and maintain safe a conditions.

Discuss the risks associated with accomplishing these actions.

SSA RAJ 02 Provide the following pertaining to non-power 0 provided in Section 4.3 and Attachment D of the LAR:

a. 'KSF pinch implementati I be done level, identify and describe

~n".rnQnt tools, and any other (KSF) identified as part of procedures such as

b. not previously been req uirements may have Ie selection was performed in accordance list of the additional components and a list functional requirement for NPO.

safe shutdown function and the NPO

'r,nT',nn by system indicating why not be included in the at-power analysis.

by fire area that were identified in the NPO fire area identification of unavailable paths in each fire area.

I be identified to the plant staff for implementation.

d. Du rious actuation of valves can have a significant impact on the ability heat removal and inventory control. Provide a description of any actio to minimize the impact of fire-induced spurious actuations on power 0 Ives (e.g., Air Operated Valves (AOVs) and Motor Operated Valves (MOVs) during PO (e.g., pre-fire rack-out, actuation of pinning valves, and isolation of air supplies).

e. During normal outage evolutions certain NPO credited equipment will have to be removed from service. Describe the types of compensatory actions that will be used during such equipment down-time.

f. The description of the NPO review for the LAR does not identify locations where KSFs are achieved via recovery actions or for which instrumentation not already included in

the at-power analysis is needed to support recovery actions required to maintain safe and stable conditions. Identify those recovery actions and instrumentation relied upon in NPO and describe how recovery action feasibility is evaluated. Include in the description whether these variables have been or will be factored into operator procedures supporting these actions.

SSARAI03 LAR Attachment 8, Table 8-2, "Plant Modifications Committed" lists the proposed modifications 82-1; 82-2; 82-3; 82-4; 82-5; 82-6; 82-7; 82-8; 82-9; 82-10; 82-11' 2; 82-13; 82-14; 82 15; 82-16; and 82-18. With respect to fire risk reduction in place, provide a statement regarding whether or not fire risk reduction measu een implemented in accordance with the plant's fire protection program for the Ii SSA RAI 04 LAR Table 8-2 describes modification E8R-11 of the fire's effect and supported by a DG in Provide a more detailed description of this mod identification of the planned physical hardware changes or l:\f1,',T'/'Inc:o ....l:\t'l.... rl~ between local control stations, the operation of the equ control process with the new modifications

Programmatic RAI 01 Based on the NRC staff's review of the LAR and during the subsequent audit, it was determined that the licensee did not adequately describe the effects of NFPA 805 on the configuration and change control processes.

Describe how the various configuration control and change control procedures are implemented together to ensure compliance with the NFPA 805 change evaluation and configuration control requirements.

Programmatic RAI 02 NFPA 805, Section 2.7.3.4, "Qualification of Users", states personnel who use and apply engineering analysis and numerical models (e.g techniques) shall be competent in that field and experienced in the app they relate to nuclear power plants, nuclear power plant fire Describe how the training program will be process, including positions that will be tra classroom, computer-based, reading program).

Programmatic RAI 03 Based on the NRC staffs review of the uent audit, it was determined that the licensee did not adequately ble loading controls and Fire Probabilistic Risk (F Describe how the assumptions r.cn<:>rn

Fire Modeling (FM) RAI 01 Section 4.5.1.2, "Fire PRA" of the Transition Report states that fire modeling was performed as part of the FPRA development (NFPA 805 Section 4.2.4.2). Reference is made to Attachment J, "Fire Modeling V&V," for a discussion of the acceptability of the fire models that were used.

Regarding the acceptability of the PRA approach, methods, and data:

1. During the audit, the technical approach for detailed fire modeling of fire compartments in support of the FPRA for the R.E. Ginna Nuclear Power Plant 805 Transition Project was discussed. The NRC staff identified the following related to this general discussion:

a. It was discussed that mechanical ventilation Consolidated Model of Fire and Smoke justification that the assumptions made

b. CFAST has been used for ca compartments, as detailed in audit enclosed obstructions was cons explain why the presence affect the results of the

c. ine sprinkler activation,

) value chosen for these RTI of the actual sprinklers

d. nt fires and transient fires due to hotwork, it or other FPRA targets are near the floor, of 15 kW is assumed as the critical fire to the postulated fire." Provide additional Ication for why this is applicable for fires close to was discussed and it was stated that, "for some fire

• the sheer size of the fire zone and openings to other zones of a hot gas layer for numerous relatively small ignition would remain localized. As such, the analysts identified those be qualitatively screened from the CFAST analysis." Provide ation about the criteria used to qualitatively screen a fire zone and those fire zones which were screened from the hot gas layer CFAST

f. Several assumptions related to fire modeling hand calculations were discussed and lead to the following questions.

i. A fire dimension of 2 ft. has been assumed for all postulated fires. Explain why this generic assumption is valid for all ignition sources across the plant.

ii. Section R.4.2.1 of NUREG/CR-6850 prescribes taking the characteristic length of the fire as equal to the cabinet's length (or vertical section as appropriate) for the purpose of calculation fire propagation through cable trays. Justify the use of a characteristic length of 2 ft. as this may not satisfy the NUREG/CR-6850 criterion.

g. Sprinkler activation in the Air Handling Room and ABM-C1 was discussed. It was stated that since sprinklers are located within the cable trays, it is reasonable and conservative to assume that, with the exception of the first cable tray above the ignition source, the remaining trays are protected by the of the sprinklers.

The activation of an automatic Halon system in discussed. It was stated that the same rationale tray above the ignition source is damaged protected, due to activation of the Halon Provide justification for this assu the sprinkler system can be applied to the Halon

h. . It was stated the model el calculation utilizes a radiative that for all the calculations the

, where the 40% radiative

. Confirm which value for the n what value was used for the value used.

i. nducted in fire compartment BR1A was lated for this analysis: a cabinet fire, secondary combustibles, a battery charger with two overhead cable trays modeled as fire, also located in the center of the room, cables Explain why two of the fires postulated in the secondary combustibles overhead and the cabinet fire (with does not. This type of discrepancy is also noted in fire

1. Provide a similar explanation for any fire area with

j. modeling analysis accounts for the potential increase in heat by the spread of a fire from the ignition source to secondary

2. Of particular concern are fires in the proximity of a wall or a corner. The entrainment of air into the flame of these types of fires is restricted compared to fires of the same size in the open. The reduced air entrainment results in higher plume and upper gas layer temperatures.

a. What are the criteria (i.e., distance from a wall or corner) that were used during the walk-downs to determine whether wall or corner effects have to be accounted for in the fire modeling analyses and provide a basis for the acceptability of the criteria.

b. Explain how wall and corner effects are accounted for in the flame height, plume temperature and ceiling jet temperature calculations.

c. Explain how wall and corner effects are accounted for in the CFAST hot gas layer calculations.

3. During the audit, the technical approach for determining the time to abandon the Main Control Room (MCR) for several fire scenarios was discussed. The NRC staff identified the following questions related to this discussion:

a. It was stated that, "Although the scenarios nr~'C1onT~ calculation were chosen as a representative set, the results of this cal be assumed to be bounding results. Additional calculations may ssess the impact of fire scenarios not specifically identified in this audit walkdown of the MCR, an office directly across from oard (MCB) was observed with a typical workstation (trash can).

Explain how a fire in this part of nrt".n ....... ent times and provide reasonable in the analysis are bounding.

b. It was stated that, "The g the main control boards, may contain [IEEE-383] qual non-IEEE-383 qualified cables (thermoplastic), or both."

cables are assumed to be to cables as defined in Table IEEE-383 qualified cables are assumed to defined in Table 8-2 of NUREG/CR 6850.

An IEEE-383 qualified cable mayor ble" as defined in NUREG/CR-6850. It is ified cable actually meets the NUREG/CR characteristics used in the 'fire modeling analysis for the and provide a basis for the assumptions above .

• nC1.'ont fires are assumed to reach the peak heat release rate However, the transient fire ramp function reaches the peak this discrepancy.

d. It was 'The panels at R.E. Ginna are closed .. ." However, it was noted during the walkdown that the interior of the MCB is a large open space. Provide technical justification for the assumption that the MCB panels are closed, and for not using the HRR distribution for Case 5 in Table E-1 of NUREG/CR-6850, Vol. 2.

e. It was discussed that a cabinet fire is assumed to propagate to adjacent cabinets in 10 minutes. In the MCR walk-down during the onsite audit staff noted that there are no internal walls between different sections of the MCB. Based on the observed field conditions, the assumptions concerning fire propagation between cabinets in the

analyses do not appear to be valid. Justify the assumptions concerning fire propagation in the MCB. Provide justification for deviating from the standard method provided in Appendix L of NUREG/CR-6B50. Perform a sensitivity analysis to assess the effect of more rapid propagation between sections of the Main Control Board. Quantify the impact on core damage frequency (CDF), ~CDF, large early release frequency (LERF) and ~LERF.

f. Several figures were discussed, which show the variation of the fire diameter with the peak heat release rate for transient and panel fires, . Each figure shows two curves, one curve corresponding to the low (0.4) and one curve corresponding to the high limit (2.4) of the NUREG- number validation range. Explain how these curves were used in d the fire diameter for the different bins of transient and panel fires. the fire area(s) were determined for the MCB fires that involve multi

g. It was stated that the abandonment scenarios is determined on the basis of the ulator (FDS) abandonment time calculations for tification for this approach as a transient fire g versus 12 min for panel files) and time compared to a panel the FM RAI 02 American Society of Mecha (ASME/ANS) Standard RA-S-200B, "Standard uency Probabilistic Risk Assessments for ", Part equires damage thresholds be established to ) must be considered in determining the potential for TnCrrn., temperature and critical heat flux criteria must be used in the

1. in block was characterized, specifically with reshold temperatures and critical heat flux for thermoset and in NUREG/CR-6B50.

2. at four cables, which run through one of the battery with thermoset damage thresholds for the purposes of fire

. Provide justification for treating these cables differently the plant.

FM RAI 03 Section 4.5.1.2, "Fire PRA" of the Transition Report states that fire modeling was performed as part of the Fire PRA development (NFPA B05, Section 4.2.4.2). Reference is made to Attachment J, "Fire Modeling V&V," for a discussion of the verification and validation (V&V) of the fire models that were used. Furthermore Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA B05" of the Transition Report states that "calculational models and numerical methods used in support of compliance with 10 CFR 50.4B(c) were verified and validated as required by Section 2.7.3.2 of NFPA B05."

1. LAR Table J-1 lists Detection Activation Model (Heat and Smoke Detection), which is not validated in NUREG-1824 (as the other referenced models in Table J-1). Table J-1 has a footnote that states that the model "is the prevailing model for estimating activation times."

This statement does not provide sufficient basis to determine the adequacy of the V&V.

Provide additional information and documentation to determine the acceptability of the model.

2. LAR Attachment J (page J-2) refers to the draft RG DG-1218 published in March 2009 for the acceptability of the fire models that were used in the application. Draft RG DG-1218 is the preliminary draft to RG 1.205 and therefore is not the ap ance. Clarify that the models used in the application are in accordance with the guidance.

3. The licensee states on page J-4 of the LAR that parameters for the CFAST files were not evaluated against the ava in NUREG-1824. It should be noted that in some calculations, with the reactor and turbine buildings, there are relatively complex covered by the V&V criteria in NUREG-1824." This ~t.:l1rQrYliQnt to determine the adequacy of the V&V. Provide ann '1rlt".,n determine the acceptability of the model, in particular validation range.

4. Based on review of Attachment to the use of FDS is provided.

However, based on discussions that FDS was used to conduct the main control room that this is the only location where FDS was used. how (what was objective) and where (fire "',"""','7 FM RAI 04 Section 4.7.3, iin Section 2.7.3 of NFPA 805," of the Transition Report and numerical models used in support of complian used as required by Section 2.7.3.3 of NFPA

1. outside the range of conditions covered by the is based. NUREG-1805, "Fire Dynamics Tools (FDTs),"

has a ns and limitations that provides guidance to the user in terms of proper and for each FDT. The general limitations of use for the algebraic equations that utilized for hand calculation was discussed. It is not clear, however, how these lim were enforced on the individual fire areas or for the multi-compartment analysis. Provide a description of how the limit of applicability was determined for each fire area.

a. Include a list of all areas, zones, transient zones and scenarios for which algebraic models were used to calculate flame height, plume temperature and point source radiation. Specify for each use whether the model was used within its range of applicability, or, justify why the model was used outside the range.

b. Include a list of areas, zones and scenarios for which algebraic models were used to calculate sprinkler, heat detector and smoke detector activation. Specify for each use whether the model was used with its range of applicability, or, justify why the model was used outside the range.

2. The range of the Froude number in the analyses for the plant calculations (with the exception of the MCR) ranges from 0.7 to 3.1, whereas the range of validation is between 0.4 and 2.4. Hence, for larger heat release rates, the Froude number will exceed the validated range. Explain why it is acceptable to exceed the validation range.

In the development of MCR abandonment times, it was ind the electric panel and transient fire areas were chosen so that the Froude in the NUREG-1824 validation range (0.4-2.4). Provide justification for in the analysis, such that the Froude number falls within the validated ra that fire areas used in the simulation are consistent with actual panel or explain why it is allowable to use a fire area inconsistent with

3. Zone models may not suitable for com width aspect ratio (e.g. the Cable Tunnel In addition, the hot gas layer temperature the hot gas layer temperature by a always used within the range of Ie room I ratio, or, if not, explain why it was

4. Identify uses, if any, of CFAST . of the model and for those cases explain how a list of areas, zones and scenarios for which development.

5. Identify uses, of applicability of the model and for those cases explain nsition Report states that fire modeling was performed as pa 805, Section 4.2.4.2). This requires that qualified fire mode "O,th'r. Furthermore, Section 4.7.3, "Compliance with Quality 2.7.3 of NFPA 805," of the Transition Report states that "Cognizant and apply engineering analysis and numerical methods in support of CFR 50.48(c) are competent and experienced as required by Section 2.7.3.4 Regarding of users of engineering analyses and numerical models (i.e., fire modeling techniques):

1. What are the licensee's requirements to qualify personnel for performing fire modeling calculations in the NFPA 805 transition?

2. What is the process for ensuring that the fire modeling personnel meet those qualifications, not only before the transition but also during and following the transition?

3. When fire modeling is performed in support of FPRA, how is proper communication between the fire modeling and Fire PRA personnel ensured?

FM RAI 06 Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805" of the Transition Report states that "Uncertainty analyses were performed as required by 2.7.3.5 of NFPA 805 and the results were considered in the context of the application. This is of particular interest in fire modeling and Fire PRA development."

Regarding the uncertainty analysis for fire modeling:

1. NFPA 805, Section 2.7.3.5 states that, when a ncrtnrr approach is used, an uncertainty analysis shall be performed to assurance that the performance criteria have been met. According 1, "Guidance on the Treatment of Uncertainties Associated cision Making,"

there are three types of uncertainty aSSClclatE

a. Parameter Uncertainty: Input statistical distributions or estimated from uncertainty of these fire modeling analysis.

in the detailed fire

b. phys mena lead to simplifying model equations. In addition, the numerical Iytical solution can lead to inexact results.

to the fact that a model is not a complete esigned to simulate. Some consider this a because most fire models neglect certain physical considered important for a given application .

. and completeness uncertainty were addressed in the

Probabilistic Risk Assessment (PRA) RAI 01 LAR page W-3 states that the variant condition includes all modifications for NFPA 805 transition, which include proposed modifications that 1) deterministically resolve variances from deterministic requirements (VFDRs), or at least reduce their delta risk, or 2) reduce plant risk but are not directly related to any particular VFDR. Clarify if "deterministically resolve VFDRs" is the same as "make the plant compliant." Provide a table with the following information: a) modifications which make the VFDR compliant, b} modifications which help to reduce delta risk but do not make the VFDR compliant, and c) modifications which reduce plant risk but are not directly related to any particular VFDR. For parts a) and b), indicate VFDRs for which it is credited and provide the technical basis, including an explanation the modification resolves, or helps to resolve the VFDR, and any relevant (e.g. design, success criteria, calculations, etc.).

PRARAI02 Different VFDRs identify different fire scenarios modification or recovery actions may be seen fire scenarios expected to be encountered, how modifications can be credited as they have been i Discuss the technical basis, including success supporting calculations, for each modification or recovery action scenarios, and cite relevant references.

PRARAI03 FR Part 21 report on the lity of the SHIELD passive thermal uring post-service testing. The likelihood Is was based on Revision 1 to rt WCAP-17100-P/NP, "PRA Model for pany (I&M) notified the NRC about recent operating indicates that the likelihood and magnitude of med from these types of seals.

s recent operating experience with the new RCP seals 21 notification have on the CDF, LERF, delta-CDF and delta for any revised assumptions.

b) If the im uses the acceptance guidelines of RG 1.174 to be exceeded, how do you plan to address this operating experience?

c) Discuss whether any currently proposed license conditions are affected by these changes.

PRARAI04 For VFDR-CC-044, discussed in the LAR, modification ES-12-0142 provides a 45 minute fire protection for cable C0687, yet this cable is not mentioned in the VFDR. Please clarify.

PRA RAJ 05 Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies that recovery actions must be addressed. Table W-4 of the LAR provides the results of the additional risk of recovery actions. However, it is not clear if both previously approved and new recovery actions are included in the results. Discuss how your treatment of previously approved and new recovery actions credited in the Fire PRA is consistent with the guidance in RG 1.205. Discuss your treatment and secondary recovery actions credited as well.

PRARAJ 06 Section 2.4.3.3 of NFPA 805 states that the PRA a acceptable to the NRC. RG 1.205 identifies that recovery actions associated with modifications, method of quantifying the human error probab the verification of assumptions made in qua implementation items in Attachment S of the LAR.

PRA RAJ 07 With respect to Frequency Asked Q and FAQ- 07-0030, discuss which method(s) were employed to eva actions from fire scenarios outside the rna' n control room abandonment, or fire the MCR and at a primary control station (PCS) he guidance for the additional risk of recovery actions Rev. 1 and the section "Additional Risk of Recovery Actions - ." Discuss how primary control stations, planned modifications were considered. Also, for the main control , describe the compliant case used. In particu nels ABEL nd IBELIP are treated for the compliant plant given y has changed to use new equipment and does not require use rding to the LAR Attachment G.

G-1, explain what is meant by "Risk" In the column "RNPCS?"

PRA RAJ 09 The LAR Attachment indicates that PCS actions are in Table G-1; however, only recovery actions appear to be in Table G-1. Please provide the PCS actions, or clarify the information in Table G-1.

PRA RAI10 Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies that recovery actions must be addressed. Explain which recovery actions and modifications are being proposed to meet risk acceptance criteria?

PRA RAI11 Describe the method(s) used to determine the fire area changes in risk (delta-CDF and delta LERF) reported in the LAR Appendix W for fire scenarios outside the main control room, fire scenarios involving main control room abandonment, or fire scenarios which may involve actions both in the MCR and at a PCS. The description should include a summary of PRA model additions or modifications needed to determine the reported changes in risk. If any of these model additions used data or methods not included in the FPRA peer review please describe the additions.

With respect to FAQ 08-00S4, Rev 1 and FAQ 07-0030, ind in the FAQs was used to evaluate the VFDRs. Include in how recovery actions, whether primary or secondary, are treated. Also, for the is noted in the LAR that recovery actions are set to success to represent a is deterministically resolved. Discuss the practice of setting recovery ivalency to removing the VFDR.

PRA RAI12 For the main control room analysis, alysis (HRA) was performed following NUREG-1921. Confirm actions was considered, consistent with NUREG-1921 gu reliability analysis. For those HEPs which were screening values, , based on the audit, it appears to be different from the main described in NUREG 1921.

PRA RAI14 Provide a discussi for plant change evaluations post transition. Include a guidance for plant change evaluations addresses analyses, and peer review facts and the PRA approach, methods, and data shall be identifies NUREG/CR-68S0 as documenting a methodology orses, with exceptions and clarifications, Nuclear Energy as providing methods acceptable to the staff for adopting a fire t with NFPA-80S.

Describe how is treated in your analysis. Address suppression with regards to protection of cable trays. Also describe your treatment of suppression with regards to progression of events.

PRA RAI16 Section 2.4.3.3 of NFPA 80S states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.20S identifies NUREG/CR-68S0 as documenting a methodology for conducting a fire PRA and endorses, with exceptions and clarifications, NEI 04-02, revision

2, as providing methods acceptable to the staff for adopting a fire protection program consistent with NFPA-805.

Transient fires should at a minimum be placed in locations within the plant PAUs where CCDPs are highest for that PAU, i.e., at "pinch points." Pinch points include locations of redundant trains or the vicinity of other potentially risk-relevant equipment, including the associated cabling.

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. Describe how transient and hot work fires are distributed within the PAUs at your In particular, identify the criteria for your plant which determine where an ignition ced within the PAUs.

Also, if you have areas where no transient or hot work fires are since those areas are considered inaccessible, define the criteria used to define "i " Note that an inaccessible area is not the same as a location where fire' , even if highly improbable.

If you have used an influence factor outside of provide a sensitivity analysis using the f"'r.r'roC' 6850. Discuss the use of the area weighting the PAUs.

PRA RAI17 A number of VFDRs involve HEMYC in the LAR involve performance-based evaluations of wra plicable, describe how wrapped or embedded mode assumptions and insights on how the PRA model bles to R delta-risk evaluations, including the model sociated with the cables.

PRA RAI18 entify FAQ 08-0053, "Kerite-FR Cable r Kerite cables are utilized at the plant, and if they describe their modeling such as the cable damage According to where water from fire suppression efforts will likely enter a potentially (e.g., a panel with unsealed penetrations or an unshielded electrical motor), to include that component in the fire scenario damage set.

Verify that the im system activation on component operation has been addressed according to Section 11.5.1.2 of NUREG/CR-6850.

PRA RAI20 RG 1.174, Revision 2, identifies that key sources of model uncertainty should be identified and sensitivity analysis performed or reasons given as to why they are not appropriate for the application. In the ASME/ANS PRA standard a source of model uncertainty is labeled as "key" when it could impact the PRA results that are being used in a decision, and consequently, may influence the decision being made.

Discuss the FPRA key sources of uncertainty and assumptions, including any related to planned modifications, and discuss the results and significance of sensitivity analyses for them.

PRA RAI21 RG 1.174, Revision 2, identifies that key sources of model uncertainty should be identified and sensitivity analysis performed or reasons given as to why they are not appropriate for the application. In the ASMEIANS PRA standard a source of model un nty is labeled as "key" when it could impact the PRA results that are being used in a consequently, may influence the decision being made.

With respect to SR FSS-E3, provide a list of the input uncertainty intervals, and their uncertainty treatment PRA RAI22 Section 2.4.3.3 of NFPA 805 states that the and data acceptable to the NRC. RG 1.205 identifies documenting a methodology for conducting a fire PRA and end with ov,r-crul rifications, NEI 04-02, revision 2, as providing methods acceptable ire protection program consistent with NFPA-805. RG 1.200 describes an associated ASME/ANS standard (currently ASMEI roach for determining the technical adequacy of the PRA once es or models have been established. ASME/AN 09 PRA should be characterized as a s regarding the internal events or fire PRA to the internal events or fire PRA subsequent to your most rOl"',o.nt, a) and fire PRAs consider the clarifications

) 1.200, Revision 2, "An Approach for Adequacy Risk Assessment Results for Risk (ADAMS Accession No. ML090410014) to the not, provide a self-assessment of the PRA model for the Ififi,,,~tin",Ct and indicate how any identified gaps were b) AS describes when changes to a PRA should be characterized as a"PRA 1.200 Revision 2 provides clarification on PRA upgrade.

Identify a made to the internal events or fire PRA subsequent to your most recent peer review. Also, address the following:

i. If any changes are characterized as a PRA upgrade, indicate if a focused-scope peer review was performed for these changes consistent with the guidance in ASME/ANS-RA-Sa-2009, and describe any findings and their resolution.

ii. If a focused-scope peer review has not been performed for changes characterized as a PRA upgrade, describe what actions will be implemented to comply with the ASME/ANS standard.

iii. If any changes for which the methodology utilized in the current fire PRA differs from that evaluated by the peer review such that a reduced capability category would result, describe what actions will be implemented to comply with the ASME/ANS standard. If this means that CC-II or greater is not met, then provide justification as to why this is acceptable for transition to NFPA 805.

PRARAI23 Section 2.2 (h) of NFPA 805 states that general approach of the standard shall involve performing the plant change evaluation that demonstrates that risk, defense-in depth, and safety margins are acceptable. Section 2.4.4.2 of NF states that the plant change evaluation shall ensure that the philosophy of is maintained, relative to fire protection and nuclear safety. If anyone of these is al fire protection features or other alternatives shall be implemented. that the plant change evaluation shall ensure that sufficient safety margins a echelons, as defined in NEI 04-02, and the general strategy of nce in the echelons is described at a high level in Section 4 to the LAR provides additional information on evaluation of following questions related to DID and SM.

a) Describe the methodology that nse-in-depth and that was used to evaluate safety marg include what was evaluated, how the evaluations were ns or changes to the plant or procedures were taken to e-in-depth or sufficient safety margins.

b) c) ,........ "'T< is a dominant contributor to LERF.

(MOV-313 and MOV 371) reliance on be placed on a single status light indication your defense-in-depth evaluation of ISLOCA. Discuss any to ISLOCA, or if none, how this conclusion satisfies Section 2.4.3.3 that the PRA approach, methods, and data shall be acceptable to the 2.4.4.1 of NFPA-805 further states that the change in public health risk arising from the current fire protection program to an NFPA-805 based program, and all future plant changes to the program, shall be acceptable to the NRC. RG 1.174 provides quantitative guidelines on core damage frequency, large early release frequency, and identifies acceptable changes to these frequencies that result from proposed changes to the plant's licensing basis and describes a general framework to determine the acceptability of risk-informed changes. Table S-3 of the LAR provides items (procedure changes, process updates, and training to affected plant personnel) that will be completed prior to the implementation of the new NFPA 805 fire protection program.

Implementation item 19 states "The following procedure change to be implemented as part of NFPA 805 transition provides a reduction in risk: A procedural change, not to be implemented until all required modifications are installed, will eliminate ER-Fire 2,3, 4, 5, and 6." Please discuss this item.

PRA RAI25 Section 2.4.3.3 of NFPA 805 states that the probabilistic safety assessment (PSA) (PSA is also referred to as PRA) approach, methods, and data shall be acceptable to the authority having jurisdiction (AHJ), which is the NRC. RG 1.205 identifies N as documenting a methodology for conducting a fire PRA and endorses, with clarifications, NEI 04-02, revision 2, as providing methods acceptable to the dopting a fire protection program consistent with NFPA-805. RG 1.200 describes a process utilizing an associated ASME/ANS standard (currently AS acceptable approach for determining the technical adequacy of approaches or models have been established. The F&Os recorded by the peer review and the <:>u,",;;:)""

Address the following questions on the dispos pporting Requirement (SR) assessment identified in that have the potential to impact the fire PRA results and do to be fully a) SC-A2. are based on the plant power uprate.

b) SY-A10. n was addressed and the c) SY-A14. the disposition was addressed and the d) events PRA key sources of uncertainty and sources of rtainty and assumptions in the FPRA, and

,.t'I"'!:lnr*", of sensitivity analyses for them.

Section 2. that the PRA approach, methods, and data shall be acceptable identifies NUREG/CR-6850 as documenting a methodology for conducting with exceptions and clarifications, NEI 04-02, revision 2, as providing m ble to the staff for adopting a fire protection program consistent with NFPA-805. RG describes a peer review process utilizing an associated ASME/ANS standard (currently E/ANS-RA-SA-2009) as one acceptable approach for determining the technical adequacy of the PRA once acceptable consensus approaches or models have been established for evaluations that could influence the regulatory decision. The primary results of a peer review are the F&Os recorded by the peer review and the subsequent resolution of these F&Os.

Address the following questions on the dispositions to the fire F&Os and SR assessment identified in Attachment V of the LAR that have the potential to impact the fire PRA results and do appear to be fully resolved.

a) CS-A 1-01. According to discussion with the licensee, Ginna contains current transformers (CTs) with a turn ratio greater than 1200:5. The phenomena identification and ranking table (PIRT) panel results, documented in NUREG/CR-7150, has not concluded that such CTs are not risk significant. Provide a sensitivity study that evaluates CDF, LERF, delta-CDF and delta-LERF for CTs of turn ratios greater than 1200:5, assuming a fire induced secondary open circuit will cause a secondary fire.

Include a description and justification of the method and associated probability applied in the quantification. VFDR-8R1 8-008 appears to be an example of this; however, it was evaluated qualitatively b) FO-F1-02. The disposition of the finding should be e consistent with Regulatory Guide 1.200, Rev. 2 which defines a sig slightly different that that noted in the disposition, i.e., 95% versus c) FSS-C1-01. Please provide a description of (the disposition cites a reference for the resolution).

d) FSS-D1-01. Please provide a descripti sposition cites a reference for the resolution).

e) FSS-D7-01. SR FSS-D7 is pression systems unavailability, including consideration of 'rn\J'lnQ a discussion on how capability category II was addre respect to meeting capability category II.

f) was related to potential fires in ube oil storage room with respect to a multi room. Given this clarification of the finding, g) in addressing the ignition frequency bins the criteria for screening CDF and LERF, but does r".nlrln with respect to the clarification in Regulatory for 0 1-01. Please discuss the results also. Please resolution of this finding (the disposition cites a reference for i) notes that not all fire areas were walked down. The disposition indicates not walked down a qualitative assessment was performed. This may not necessarily be sufficient such as for fire areas with a potentially significant CCDP or CLERP. Since the focus of supporting requirement SF-A 1 is on initiating events, identifying unique fire ignition source scenarios may be important. For fire areas which have not been walked down and are potential risk Significant in the Fire PRA, either perform a walkdown or provide in further detail the justification for determining that unique sources do not exist or are already known for such fire areas.

j) SF-A2-01. Please provide addition information discussing the reviews and assessments performed to meet supporting requirement SF-A2.

PRARAI27 Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a fire PRA and endorses, with exceptions and clarifications, NEI 04-02, revision 2, as providing methods acceptable to the staff for adopting a fire protection program consistent with NFPA-805. In letter dated July 12,2006, to NEI, the NRC established the ongoing FAQ process where official agency positions regarding acceptable methods can be documented until they can be included in revisions to RG 1.205 or NEI 04-02. Methods that have not been' determined to be acceptable by the NRC Staff require additional n to allow the NRC Staff to complete its review of the proposed method.

Identify and describe all UAMs or deviations from NU approved FAQs, and clarify whether guidance from the J "Recent Fire PRA Methods review Panel Decisions Heat Release Rates in Electrical Cabinets Fires'"

identified deviations from NUREG/CR-6850 that sensitivity study that estimates the impact of delta-CDF, and delta-LERF.

PRA RAI28 It was recently stated at the industry na Identification and Ranking Table Panel being conducted for the SIREE-FIRE and CAROL FIRE tests may be eliminating the credit (CPTs) (about a factor 2 reduction) currently a les 1 850, Vol. 2, as being invalid when CPT credit reduction is included in the Fire e'O"Tr"",,',,, as described in NUREG/CR-6850 outside of cabinets, within and without the rt""'O"t screening based on the hot gas layer ment. Include a description of how identification of ed. If the impact of fire on sensitive electronics risk not performed, provide the contribution of these RF, and delta-LERF using recommended criteria from sensitive electronics.

PRARAI30 During the audit, it was observed that several bus ducts run through the relay room. Provide a discussion of the analysis which describes the frequency and zone of influence (ZOI) associated with bus duct fires in the relay room and compare to FAQ-07-0035 in NUREG/CR-6850 Supplement 1. Indicate if the ZOI from the FAQ extends beyond the single transient zone.

PRA RAI31 Discuss the transient zone approach for fire damage. Discuss if scenarios across transient zones were identified and evaluated in the PRA. Discuss the mechanism for propagation

across transient zone boundaries. List the scenarios that cross the boundary that were included in the PRA, identifying the originating and final transient zones. Provide the contribution of these scenarios to CDF, delta-CDF, LERF, and delta-LERF.

PRA RAI32 Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a fire PRA and endorses, with exceptions and clarifications, NEI 04-02, revision 2, as providing methods acceptable to the staff for adopting a fire n program consistent with NFPA-805. The ASME/ANS PRA standard includes for seismic-fire interactions are a part of the Fire PRA review due to the f this issue in the ASME/ANS PRA standard. The seismicities for central and were changed as a result of the USGS re-evaluation (USGS, "2008 NSHM Gri, Ground Acceleration"), based on reanalysis of the New Madrid if your evaluation of seismic-fire interactions includes the results of the how the USGS re-evaluation was addressed.

PRA RAI33 The main control board fire scenario mot....."'nc and justification ASMEIANS PRA standard with respect to NU se provide additional information on the following. M.>JIVIt;;~/M.1 a) Appendix L for the bility of damage which bility, and the separation analysis uses the fire ignition frequency 08-0048, Section 10.2.1), the approach L. Describe the approach taken in lieu of re L-1), and provide justification for its resses the considerations which were pplicability to the MCB analysis CR risk analysis of the spread of fire between panels I£>T"",:::u:::.n panels and how it is consistent with NUREG/CR justification for its use.

The peer review a Applicable to SRs PRM-B3, PRM-B4, PRM-B6, PRM-B8, PRM B9, PRM-B10, and 15. Please provide the basis for this assignment.

PRARAI35 For SR SF-A5 it appears that the normal expected practice of fire-fighting personnel is credited to meet this supporting requirement. With respect to fire brigade training procedures related to earthquakes, it was confirmed at the audit that these procedures do not exist. Normal expected practice of fire-fighting personnel may not be sufficient for seismic-fire scenarios. Please assess the seismic-fire scenarios against the normal expected practice and discuss the

conclusion on the need to develop training procedures which account for potential earthquake fire-related challenges.

PRARAI36 Fire-induced instrument failure should be addressed in the HRA per NUREG/CR-6850 and NUREG-1921, and instrumentation credited for operator actions in the HRA should be verified to be available. Therefore, please address the following:

a) Describe how fire-induced instrument failure (including no off-scale readings, and incorrect/misleading readings) is addressed in the fire Include discussion of instrumentation that was modeled explicitly in the fault success criteria assumed for this modeling, and how explicit model norU<:>I'.nn was done in the evaluation of HEPs.

b) Confirm that instrumentation credited in the scenarios in which they are credited.

PRARAl37 During the audit it was noted that the truncated at a floor due to independence considerations. re difficult to justify and a floor is reasonable. NUREG-1792 notes tha lities of aI/ the human failure events in the same accident a justified value, and suggests an HEP floor of 1E-5. using this probability as a floor. Provide the LAR Ta LERF results.

PRARAI38 SRs in the ASMEI updating ignition frequencies if outliers are found (SR IGN-A4) or equivalent statistical process (SR IGN A6). The fi updating was not performed.

Please IJCI'vl;;;:,ian updating with plant-specific data.

Supp ent 1 states that a sensitivity analysis should be ition frequencies in the Supplement instead of the fire ignition frequencies 1 of NUREG/CR-6850. If the frequencies from FAQ 08-0048 were used, was nsitivity analysis performed? If not, provide the sensitivity analysis of the im the Supplement 1 frequencies instead of the Table 6-1 frequencies on CDF, , delta-CDF, and delta-LERF for all of those bins that are characterized by an that is less than or equal to one. Perform this sensitivity analysis using the baseline established by PRA RAI45 for CDF, LERF, delta-CDF, delta-LERF. Provide the results, and if risk acceptance criteria are exceeded, consider fire protection, or related measures that can be taken to provide additional defense-in-depth.

PRARAI40 Discuss your plans for modeling the charging pump motor vari-drive replacement, including the location of the variable drives, in the Fire PRA, and whether or not this change is part of Implementation Item 9 in the LAR, Attachment S.

PRARAI41 The peer review notes that the PRA assumes staggered testing for all components subject to CCF. However, main steam isolation valves (MSIVs) are apparentl tested on a staggered basis. Please confirm that you have addressed this observation Vs. If not, please discuss your plans, and whether or not this change is part of n /tem 9 in the LAR, Attachment S.

PRARAI42 The INTAKE structure was qualitative screened only cable self ignition. If the INTAKE structure does not meet be included in the Fire PRA since screening based the criteria. In addition, it appears that only self-ig postulated for cables above water, while SR IGN-A9 requires the IJV<>lUICl combustibles for physical analysis units, so that this may also e Fire PRA.

PRARAI43 NUREG/CR-6850 d ng step for

.multi-compartments. for the multi-compartment analysis. all the approach and criteria is consistent with NUREG/CR-6850. an explanation. The screening analysis ensure an adequate level of screening for of 1.0E-9. Provide the results of re-Tab ry," of the LAR provides the risks associated with the VFDRs. LERF values for a number of fire areas such as EDG1A, , SH, STA13ACH, and YARD to be "O.OOE+OO." As these fire areas is clear how these areas can have zero contribution to LERF and delta-LE none of the fire-induced core damage sequences in these fire areas lead to sequences (justify the zero value contributions), provide the actual calculated va these fire areas, or explain what these values are meant to represent.

PRARAI45 Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a fire PRA and endorses, with exceptions and clarifications, NEI 04-02, revision 2, as providing methods acceptable to the staff for adopting a fire protection program consistent with NFPA-805. The Fire PRA model should reflect the post-transition modifications and the as

built plant, and incorporate acceptable methods. To address these considerations, make the following changes to the Fire PRA model to redefine the baseline of your Fire PRA using acceptable methods.

• changes from RAI 3 involving the planned reactor coolant pump seals;

• changes from RAI 26 a) involving current transformers;

• changes from RAI 29 involving sensitive electronics;

• changes from RAI 31 involving transient zone fire modeling;

• changes from RAI 37 involving the cutset HEP floor;

• changes from RAI 43 involving multicompartment analysis Provide the LAR Table W-4 CDF, LERF, delta-CDF, and