ULNRC-06011, Responses to RAI Set 3 Regarding Adoption of National Fire Protection Association Standard NFPA 805

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Responses to RAI Set #3 Regarding Adoption of National Fire Protection Association Standard NFPA 805
ML13218A171
Person / Time
Site:  Ameren icon.png
Issue date: 08/05/2013
From: Maglio S
Ameren Missouri
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
ULNRC-06011
Download: ML13218A171 (33)
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Text

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WAmeren Callaway Plant MISSOURI August 5, 2013 ULNRC-060 11 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 10 CFR 50.90 Ladies and Gentlemen:

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

FACILITY OPERATING LICENSE NPF-30 RESPONSES TO RAI SET #3 REGARDING ADOPTION OF NATIONAL FIRE PROTECTION ASSOCIATION STANDARD NFPA 805 Reference 1): Ameren Missouri Letter ULNRC-05851, Response to Request for Additional Information Regarding Adoption ofNational Fire Protection Association Standard NFPA 805 dated April17, 2012 (TAC No. ME7046)

Reference 2): Ameren Missouri Letter ULNRC-05876, Response to Request for Additional Information Regarding Adoption ofNational Fire Protection Association Standard NFPA 805 dated July 12, 2012 (TAC No. ME7046)

Reference 3): Ameren Missouri Letter ULNRC-05952, Response to Request for Additional Information Regarding Adoption ofNational Fire Protection Association Standard NFPA 805 dated February 19, 2013 (TAC No. ME7046)

On March 2, 2012, Ameren Missouri received a Request for Additional Information (RAI)

(ML120600186) related to the license amendment request (LAR) to adopt National Fire Protection Association Standard 805 (NFP A 805). Reference 1 transmitted the responses to a portion of the RAI questions.

On April 12, 17, & June 21, 2012, teleconferences between Ameren Missouri and the NRC staff reviewers were conducted in which additional questions on the remaining RAI questions were discussed. Following these teleconferences, the NRC provided supplemental questions to the original

PO Box 620 Fulton, MD 65251 AmerenMissouri.com

ULNRC-06011 August 5, 2013 Page2 (ML12159A111, ML12159A103, ML121710439, and ML121710444). Reference 2 transmitted the responses to the supplemental questions identified during the teleconferences.

On December 11, 2012 Ameren Missouri received a second RAI (ML12335A232) related to the amendment request. Reference 3 transmitted the responses to the RAI questions.

On July 30,2013 Ameren Missouri received a third RAI (ML13204A223) related to the amendment request. Enclosure 1 contains Ameren Missouri's responses to the individual requests contained in the July 30, 2013 RAI. The Enclosure also contains a description of changes to the LAR that Ameren Missouri has identified as being warranted. Attachments are provided within the enclosure to show the LAR Transition Report changes from the original submittal.

It should be noted that this letter does contain new commitments. Attachment 1 references commitments which will be implemented following NRC approval of the subject LAR. Attachment S provides details on those commitments. Only affected commitments are included in Attachment S of this letter.

Licensee Identified Change (LIC) 21 requests the time period for completion of the implementation items in Table S-3 be extended from six months to eight months after NRC approval. Refer to Section 7 of this report for details.

Ifthere are any questions with regard to these RAI responses, please contact Mr. Scott Maglio at (573) 676-8719 or Mr. Justin Hiller at (314) 225-1141.

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

Sincerely, Executed on: ----------------

~to cry Scott Maglio Regulatory Affairs Manager GAHa/nls , List of Commitments , Request for Additional Information (RAI) with Callaway Plant Response

ULNRC-060 11 August 5, 2013 Page 3 cc: U.S. Nuclear Regulatory Commission (Original and 1 copy)

Attn: Document Control Desk Washington, DC 20555-0001 Mr. Steven Reynolds Regional Administrator U.S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. Fred Lyon Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-8B 1 Washington, DC 20555-2738

ULNRC-060 11 August 5, 2013 Page4 Index and send hardcopy to QA File A160.0761 Hardcopy:

Certrec Corporation 4150 International Plaza Suite 820 Fort Worth, TX 76109 (Certrec receives ALL attachments as long as they are non-safeguards and may be publicly disclosed.)

Electronic distribution for the following can be made via Tech Spec ULNRC Distribution:

A. C. Heflin F. M. Diya C. 0. Reasoner III B. L. Cox L. H. Graessle J. S. Geyer S. A. Maglio Corporate Communications NSRB Secretary T. B. Elwood R. C. Wink J. W. Hiller G. A. Harris M. K. Fletcher L. E. Eitel S. G. Cantrell STARS Regulatory Mfairs Mr. John O'Neill (Pillsbury Winthrop Shaw Pittman LLP)

Missouri Public Service Commission Ms. Leanne Tippett Mosby (DNR)

ULNRC-060 11 August 5, 2013 Page 5 ATTACHMENT 1 LIST OF COMMITMENTS The following table identifies those actions committed to by Ameren Missouri in this document. Any other statements in this document are provided for information purposes and are not considered commitments. Please direct questions regarding these commitments to Justin Hiller (Supervising Engineer, Risk Management) at 314-225-1141.

COMMITMENT Due Date/Event COMN Callaway will complete activities necessary to June 30, 2013 50143-support the new licensing basis as indicated in 50145 Callaway Plant's NFPA 805 Transition Report Table S-2, "Plant Modifications Committed."

Callaway will complete activities necessary to The later of 6 50146-support the new licensing basis as indicated in months after 50207, Callaway Plant's NFPA 805 Transition Report Refuel Outage 19 50212, 50213, Table S-3, "Implementation Items." or 8 months* 50245, 50243, This letter includes seven new commitments. after NRC 50257,50258 approval. New commitments:

50379-50385

  • Implementation period changed from 6 months to 8 months after NRC approval. Refer to Licensee Identified Change 21 in Section 7 of Enclosure 1.

to ULNRC-06011 Enclosure 1, Request for Additional Information (RAI) with Callaway Plant Response Section 1: Response to Fire Modeling RAis Section 2: Response to Fire Protection Engineering RAis Section 3: Response to Programmatic RAis Section 4: Response to Safe Shutdown RAis Section 5: Response to Probabilistic Risk Assessment (PRA) RAis Section 6: Response to Radiation Release RAis Section 7: Licensee Identified Changes to the Transition Report : Revisions to the Transition Report Main Body Attachment A: Revisions to Transition Report Attachment A - NEI 04-02 Table B Transition of Fundamental Fire Protection Program and Design Elements Attachment B: Not used.

Attachment C: Not used.

Attachment D: Not used.

Attachment E: Not used.

Attachment F: Not used.

Attachment G: Not used.

Attachment H: Revisions to Transition Report Attachment H - NFPA 805 Frequently Asked Question Summary Table : Not used.

Attachment J: Not used.

Attachment K: Not used.

Page 1 of28 to ULNRC-060 11 Attachment L: Not used.

Attachment M: Not used.

Attachment N: Not used. : Not used.

Attachment P: Not used.

Attachment Q: Not used.

Attachment R: Not used.

AttachmentS: Revisions to Transition Report AttachmentS- Plant Modifications and Items to be completed during Implementation Attachment T: Not used.

Attachment U: Not used.

Attachment V: Not used.

Attachment W: Not used.

Attachment X: Not used.

Page 2 of28 to ULNRC-060 11 Section 1: Response to Fire Modeling RAis No new RAis were submitted for this topic.

Page 3 of28 to ULNRC-060 11 Section 2: Response to Fire Protection Ene:ineerine: RAis Fire Protection Engineering RAI 18 The compliance statement for LAR Table B-1, Element 3.4.1 (c) [On-Site Fire-Fighting Capability] is "complies".

Describe how the requirements of NFPA 805 Section 3.4 .1 (c) are met, specifically "the brigade leader and at least two brigade members shall have sufficient training and knowledge of nuclear safety systems to understand the effects of fire and fire suppressants on nuclear safety performance criteria".

An approach acceptable to the staff for meeting this training and knowledge requirement is provided in Regulatory Guide 1.189, Revision 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 possession of an operator's license or equivalent knowledge of plant systems."

Response to Fire Protection Engineering RAI 18 As indicated in LAR Table B-1, Callaway complies with the requirement stated in NFP A 805 Section 3.4.1 (c) which specifically requires that "the brigade leader and at least two brigade members shall have sufficient training and knowledge of nuclear safety systems to understand the effects of fire and fire suppressants on nuclear safety performance criteria".

At Callaway, the Fire Brigade Leader (Incident Commander) position is filled by either an Operating Supervisor or Operations Technician. The Fire Brigade Leader may be a Licensed Operator or a Non-Licensed Operator who is fully qualified on all Operations Technician watch stations.

In addition, two of the four fire brigade members are Operations Technicians (non-licensed operators) who at a minimum have completed class room training on the Primary and Secondary Operations watch stations.

The Operations Technician Training program for the Primary and Secondary Operations watch stations includes review of nuclear safety systems and objectives including understanding the purpose, major system flow-paths, operation of major system components, indications, controls, and system operation during normal and off-normal conditions. This includes discussion of components and or flow paths affected by Technical Specifications for the systems.

The Fire Brigade Training program provided to all fire brigade members includes instruction on identification and location of safety related equipment which may be affected by a fire in the area, plant systems that need to be protected from fire, fire safe shutdown (nuclear safety performance Page 4 of28 to ULNRC-06011 criteria), and safety related heat sensitive equipment. Training is also provided on the suppressant and their effects.

Transition Report Table B-1 Section 3 .4.1 (c) is revised to provide additional details of the qualifications in the Compliance Basis. The changes to Transition Report Attachment A are shown in Attachment A to this Enclosure.

Fire Protection Engineering RAI 19 Does the site have any storm drains in the yard area that discharge to the unrestricted area? If the answer is yes, then please confirm that you have completed a liquid release evaluation.

Response to Fire Protection Engineering RAI 19 For the locations and buildings identified within the YD-1 fire area, if floor drains are provided within the building they route to the sanitary sewer system where effluent is processed and then transferred to a settling pond. The settling pond supernatant is recycled to the water treatment plant or discharged in accordance with applicable permits and all processing and ponds are located within the site boundary. The sanitary sewer system is periodically sampled for radioactivity. In the event of a site fire involving radioactive/contaminated materials, Radiation Protection personnel would be involved with the fire response. HTP-ZZ-05006, Fire Involving Radioactive Material or Entry into the Radiological Controlled Area, requires follow up surveys as necessary to identify contamination spread due to firewater runoff, so there is reasonable assurance sampling and containment of contaminated effluent will occur.

For general yard satellite radioactive material (RAM) storage locations where temporary or portable Radwaste containers are located, water surcharge will drain to local storm drains or ditches which drain to site retention ponds. In the event of a fire in the YD-1 area involving radioactive/contaminated materials, Radiation Protection personnel would be involved with the fire response. HTP-ZZ-05006, Fire Involving Radioactive Material or Entry into the Radiological Controlled Area, requires follow up surveys as necessary to identify contamination spread due to firewater runoff, so there is reasonable assurance sampling and containment of contaminated effluent will occur. The spill would then be addressed by the actions specified in HDP-ZZ-07000, Radiological Monitoring Program and Groundwater Protection Initiative, and RP-DTI-ENVIRONMENTAL-SPILLRESP, Response to Spills or Leaks of Radioactive Material into Groundwater, that specifies additional containment, monitoring, dose evaluations, and reporting actions beyond those taken during the immediate fire event as called out in the pre-fire plan. All retention ponds are within the site boundary therefore there is reasonable assurance no liquid effluent would be released to unrestricted areas.

Page 5 of28 to ULNRC-060 11 Section 3: Response to Pro2rammatic RAis No new RAis were submitted for this topic.

Page 6 of28 to ULNRC-06011 Section 4: Response to Safe Shutdown RAis No new RAis were submitted for this topic.

Page 7 of28 to ULNRC-060 11 Section 5: Response to Probabilistic Risk Assessment (PRA)

RAis

Background:

The NFPA-805 standard incorporated by reference into 10 CFR 50.48(c) 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. RG 1.200 describes a peer review process utilizing an associated ASME/ANS standard as one acceptable approach for determining the technical adequacy of the PRA. In its letter to NEI dated July 12, 2006, the NRC established the ongoing frequently asked question (F AQ) 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.

NFP A-805 also requires that any change in public health risk that results from transition to an NFPA-805 based program from the plant's current fire protection program, and all future changes to the NFPA-805 based program, be acceptable to the NRC. RG 1.174 provides quantitative guidelines on core damage frequency and large early release frequency, and identifies acceptable changes to these frequencies that result from proposed changes to the plant's licensing basis. RG 1.174 also describes a general framework to determine the acceptability of risk-informed changes.

Probabilistic Risk Assessment RAI 36, Timing for Post-Fire Human Failure Events The licensee reported relatively small error probabilities for some rapid actions. Acceptable methodologies for human error probability estimates generally assign large error probabilities for rapidly required responses. The licensee's response to RAI 07-B and RAI 35 discussed three specific cases where the time "margins" for completion of critical tasks were very short (approximately one minute or less). These were dispositioned via sensitivity evaluations where each human error probability was assigned a value of 1.0 (totally unsuccessful). The reported increases in CDF, LERF, delta-CDF and delta-LERF ranged from about 3 percent to about 15 percent, remaining below the numerical acceptance thresholds in RG 1.174, as cited in RG 1.205.

Please indicate whether these sensitivity evaluations (using HEP = 1.0) will be incorporated directly into the PRA or if alternative analyses (e.g., demonstrating more substantial time margins to support the original values) will be performed for the PRA. If it is the latter, describe these analyses and provide the results.

Response to Probabilistic Risk Assessment RAI 36 Callaway will utilize alternate analyses specifically by modifying the PRA model to provide detailed HRA of all actions required after Control Room evacuation. This evaluation will be accomplished Page 8 of28 to ULNRC-06011 using the EPRI HRA calculator and be conducted in accordance with the guidance in NUREG-1921 prior to use of the Fire PRA for evaluation of post-transition changes to the Fire Protection Program.

The results of this evaluation (as requested in the RAI) will not be available until after transition to NFPA 805. The changes to Transition Report AttachmentS are shown in AttachmentS to this Enclosure.

Probabilistic Risk Assessment RAI 37, Use of Fractional Influence Factors for Transient Fires NUREG/CR-6850 provides a method for apportioning transient fire frequencies among fire areas. FAQ 12-0064 provides an alternative method. In the responses to RAI 08-A and RAI 32, the licensee reported a "deviation from NUREG/CR-6850 (EPRI 1011989)" where fractional (<1.0) influence factors were assumed for certain transient fire scenarios. A special weighting factor of 0.05 was used for Maintenance in hot work prohibited zones and a factor of0.1 was used for Storage in transient combustible free zones. Except for the reactor coolant pump (RCP) room in Containment, the minimum value for occupancy was 1.0; thus, the combined weighting factors were always greater than 1.0. The licensee indicated that these fractional values were always combined with at least a weight of 1.0 for the Occupancy influence factor. Therefore, the analyses as performed already constituted a sensitivity evaluation. However, with the publication ofFAQ 12-0064, the accepted method now restricts use of fractional values to specific cases, which are not currently represented in the licensee analysis. Also, the RCP room was assigned a 0.0 for all three factors. Personnel occupancy and maintenance work does not occur in this area during power operation, due to health and safety concerns. The final transient frequency is 0.0, a value that is inconsistent with the ASME/ANS PRA Standard, although the impact on risk is negligible.

Please re-evaluate all cases, and provide the results, where fractional values were employed, including the RCP room where a total influence factor of zero was assigned, in accordance with FAQ 12-0064, or provide justification for not using FAQ 12-0064. Please indicate what method the licensee intends to use in its PRA when estimating the change in risk associated with post-transition changes to the fire protection program. If the method is not consistent with one of the two acceptable methods, please provide a justification for the proposed method.

Response to Probabilistic Risk Assessment RAI 37 Callaway will revise its Fire PRA analysis Fractional Influence Factors for Transient Fires to those specified within FAQ 12-0064. Specifically, Callaway will revise its Fire PRA analysis to utilize the Fractional Influence Factors (e.g., 0.0, 0.1, 0.3, 1.0, 3.0, 10, or 50) in FAQ 12-0064 and ensure that all evaluated locations have a total influence factor greater than 0.0 prior to use of the Fire PRA for post-transition changes. Transition Report Table S-3 is revised to add a commitment to implement the described change prior to use of the Fire PRA for post-transition changes. The changes to Transition Report Attachment H are shown in Attachment H to this enclosure. The changes to Transition Report Attachment S are shown in Attachment S to this Enclosure.

Page 9 of28 to ULNRC-060 11 Probabilistic Risk Assessment RAI 38, Low Ignition Frequency for Bus Duct Fires The fire frequencies contained in NUREG/CR-6850 are average frequencies that have been compiled from industry wide experience and were developed to be generically applicable to each individual plant. The licensee's response to RAI 08-B cited plant-specific presence of"considerably fewer iso-phase bus ducts than a typical plant," to reduce the generic bus duct fire frequency by a factor of five. The licensee also provided the results of a sensitivity study without this reduction, which showed about 10 percent increases in CDF and LERF for the ignition frequency bin, but less than 1 percent increases in total fire CDF and LERF. There was no change in the corresponding change in risk values.

Please indicate the values the licensee intends to use in its PRA when estimating the change in risk associated with post-transition changes to the fire protection program. If the licensee proposes to use the reduced value, please evaluate whether there is any equipment that might be more common at Callaway Plant than at an average plant such that the associated frequencies should be increased, and justify why it is acceptable for the licensee to modify frequencies that have been developed to be nominally applicable to all plants.

Response to Probabilistic Risk Assessment RAI 3 8 Callaway will revise the Fire PRA to utilize the total ignition frequency for Bus Duct fires (BIN 16.2) which are specified in Supplement 1 to NUREG/CR-6850 prior to use of the FPRA for evaluation of post-transition changes to the Fire Protection Program. The changes to Transition Report Attachment S are shown in Attachment S to this Enclosure.

Probabilistic Risk Assessment RAI 39, Credit for Control Power Transformers (CPTs) for AC Circuit Failure Probabilities Based on recent developments from cable fire tests, consensus between the nuclear industry and NRC is that the current credit for reducing "hot short" probabilities when CPTs are present now appears unverifiable. Volume 1 ofNUREG/CR-7150, published in October 2012, states that, "Ultimately, the PIRT panel concluded that CPT size alone, nor indeed the mere presence of a CPT as the powering device, is not a predictable and repeatable circuit design parameter that reliably yields fewer spurious operations." The licensee's application credited the presence of CPTs to reduce the hot short probability. In response to RAI 09-A, the licensee provided the results of a sensitivity study without taking credit for the presence of a CPT (nominally a reduction in hot short probability by a factor of two). The licensee reported increases in CDF, LERF, delta-CDF and delta-LERF ranging from about 30 percent to nearly 100 percent.

Please indicate whether the sensitivity analysis will be incorporated directly into the PRA, or if the recently published guidance based on the updated spurious actuation probabilities and durations will be used. Please indicate the values the licensee intends to use in its PRA when estimating the change in risk associated with post-transition changes to the fire protection program. If the factor of two will Page 10 of28 to ULNRC-060 11 be maintained, please provide a justification for why the NUREG/CR-7150 conclusion is not applicable to Callaway Plant.

Response to Probabilistic Risk Assessment RAI 39 Callaway will revise the Fire PRA to utilize the specific CPT treatment guidance provided in a Letter to Joseph G. Giitter from Richard P. Correia dated June 14th (ML13165A209), 2013 titled "Interim Technical Guidance on Fire Induced Circuit Failure Mode Likelihood Analysis" prior to use of the FPRA for evaluation of post-transition changes to the Fire Protection Program. The changes to Transition Report Attachment S are shown in Attachment S to this Enclosure.

Probabilistic Risk Assessment RAI 40, Fire Growth Time to Peak Heat Release Rate for Trash Fires FAQ 08-0052, in Supplement 1 to NUREG/CR-6850 (EPRI 1011989), suggested the 8-min growth time for common trash fires contained within receptacles. The licensee used a 10 minute growth time. In response to RAI 10, the licensee provided the results of a sensitivity evaluation using an 8-min fire growth time as the basis instead of 10 minutes indicating only about 0.1 percent increase in the CDF for control room fires. In the RAI response, the licensee also cited a re-evaluation of the data supporting the FAQ to support its use of 10 minutes, which the NRC staff did not accept as adequate justification for deviation from the FAQ. The FAQ recommendation is based on Tests 5 and 7 through 9 ofNUREG/CR-4860 (which is the reference cited by the licensee as its basis for assuming a 10-min growth time) and NIST and LBL tests. The staff questioned the reason for the licensee's inclusion of Tests 3 and 4 from this reference in its basis, since these two tests were previously discounted when the FAQ was developed, and the licensee provided no new justification for including the tests.

Please indicate what values the licensee intends to use in its PRA when estimating the change in risk associated with post-transition changes to the fire protection program. If the licensee proposes to use the 10-minute factor, please provide additional justification.

Response to Probabilistic Risk Assessment RAI 40 Callaway will revise the Fire PRA to utilize the fire growth time to peak heat release rate for trash fires of 8 minutes specified in FAQ 08-0052 incorporated in Supplement 1 to NUREG/CR-6850 (EPRI 10 11989) for the Main Control Room transient fires prior to use of the Fire PRA for evaluation of post-transition changes to the Fire Protection Program. The changes to Transition Report Attachment S are shown in Attachment S to this Enclosure.

Page 11 of28 to ULNRC-060 11 Probabilistic Risk Assessment RAJ 41, Uncertainty Analysis for Ignition Frequencies beyond FAQ 08-0048 FAQ 08-0048, in Supplement 1 to NUREG/CR-6850 (EPR11011989), requires limited sensitivity analyses for selected ignition frequency bins while the PRA standard requires uncertainty analyses for key assumptions. The licensee's application did not include any uncertainty analyses. In response to RAI 09-B, the licensee performed a sensitivity study of all ignition frequency bins by applying a multiplication factor that ratioed the 95th percentile frequency to the mean frequency for each bin in Supplement 1 to NUREG/CR-6850 (EPR11011989).

Please confirm that this uncertainty/sensitivity evaluation will be used as the uncertainty analysis regarding fire ignition frequencies for the PRA when the licensee estimates the change in risk associated with post-transition changes to the fire protection program.

Response to Probabilistic Risk Assessment RAI 41 Callaway will implement the uncertainty methodology used in response to PRA RAI 09-B to estimate the change in risk associated with post-transition changes to the Fire Protection Program with the understanding that the uncertainty methodology can be refined to utilize parametric data evaluations.

The changes to Transition Report Attachment S are shown in Attachment S to this Enclosure.

Probabilistic Risk Assessment RAJ 42, Effect of Internal Events PRA Update of Common Cause Failures (CCFs) on Fire PRA The results of the most recent focused-scope peer review of the latest update to the internal events PRA indicated that the some CCFs were not modeled and that other CCF probabilities should be updated. The licensee's application stated that some CCFs were updated/added, but not in the fire PRA. In response to RAI 01-C and RAI 33, the licensee performed a sensitivity evaluation on CDF (CDF is the limiting risk metric at Callaway Plant, because LERF is always lower than 10 percent of CDF) using updated CCF probabilities from the internal events PRA in the fire PRA. The sensitivity analysis was conducted in two parts. The first part evaluated the CDF increase for those basic events already modeled in the fire PRA. For all the CCF events that have a direct match between the internal events and fire PRAs, the net change in both CDF and delta-CDF was negative. The second part was to evaluate the CDF increase for those basic events that are not modeled in the fire PRA. The only set of CCF events in the current internal events PRA which are not included in the FPRA are the CCF combinations of the non-safety auxiliary feed water (NSAFP) pump and the safety-related motor-driven auxiliary feedwater (AFW) pumps. For this sensitivity study, a bounding risk approach employed surrogate events (NSAFP test and maintenance events), assuming the non-safety auxiliary feedwater pump is failed (basic event probability is set to 1.0). The results indicate that the increase in CDF remains under the RG 1.205 acceptance value of 1E-5/yr.

When the licensee estimates the change in risk associated with post-transition changes to the fire protection program, please indicate whether (1) the sensitivity analysis will be incorporated directly into the PRA, or (2) the PRA will be updated to include the latest CCFs. If the licensee proposes a Page 12 of28 to ULNRC-06011 different alternative, please justify why the most recent available data and most comprehensive modeling is not needed to support self-approval.

Response to Probabilistic Risk Assessment RAI 42 Callaway will revise the Fire PRA (FPRA) to incorporate the Internal Events PRA Common Cause Failures (CCFs) such that both the current IE and FPRA are consistent regarding CCFs prior to using the FPRA to evaluate the change in risk associated with post-transition changes to the Fire Protection Program. The changes to Transition Report Attachment S are shown in Attachment S to this Enclosure.

Probabilistic Risk Assessment RAI 43, Longer than Expected Time Available to Isolate Reactor Coolant System (RCS) Injection The licensee credits about 36 minutes (min) as available to isolate the RCS injection flow to avoid challenging the PORV during pressurizer overfill. This differs from the Callaway FSAR, Section 15.5.1.2, which states that the pressurizer becomes water solid following a spurious Safety Injection signal within 9 min, even if the operator terminates normal charging pump flow at 6 min. In its response to RAI-12, the licensee cited plant-specific calculations as the basis to justify the 36-min time frame and described the scenario in detail, providing the results of the MAAP analysis, which yields 36 min. This was further compared to the RETRAN analysis used for the FSAR estimate of about 9 min. The MAAP analysis, which is appropriate for PRA, involves best estimates, whereas the RETRAN analysis involves the much more conservative design basis. The key driver among the different parameter assumptions yielding the large difference in available time is the nominal flow rate into the RCS. In RETRAN, this is conservatively assumed to be 346 gallons per minute (gpm) for 6 min, followed by 299 gpm afterward. In MAAP, a supposedly more realistic 126 gpm flow rate is assumed throughout. However, each centrifugal charging pump is capable of discharging about 125 gpm into the RCS at full reactor pressure (based on the pump curve provided in the FSAR). The MAAP run appears to only assume that one centrifugal charging pump spuriously started at time zero and that the operators trip the normal charging pump early (5-6 minutes). The RETRAN calculation reflects what would be expected: for the first 6 minutes, the injection flow rate is based on the normal charging pump (about 47 gpm) plus two centrifugal charging pumps (about 299 gpm). After 6 minutes, the injection flow rate is based on the two centrifugal charging pumps (about 299 gpm). The RETRAN model appears optimistic in the flow rate (it credits about 150 gpm per pump versus an expected 125- 130 gpm per pump). Regarding the apparently low injection flow rate in MAAP, doubling the injection flow (as would be expected for two HHSI pumps) reduces the time in half, which means that only 18 minutes rather than the reported 36 minutes would be available. In addition, the licensee performed a sensitivity analysis taking no credit for an operator recovery action, which indicated increases in CDF, LERF, delta-CDF and delta-LERF <10 percent.

When the licensee estimates the change in risk associated with post-transition changes to the fire protection program, please confirm that the sensitivity analysis will be used as the basis for the evaluation supporting the PRA, including any changes to the PRA, as needed, to reflect the sensitivity Page 13 of28 to ULNRC-060 11 assumptions and results. Otherwise, please explain the apparent discrepancy between the RETRAN and MAAP results discussed above.

Response to Probabilistic Risk Assessment RAI 43 This response supplements the Callaway Plant response to PRA RAI 12 which provided a comparison of the inputs and assumptions contained in the Design Basis Accident Analysis described within FSAR Section 15.5.1.2 and the assumptions and results of the thermal-hydraulic (T-H) analysis utilized to determine the feasibility of recovery actions for a fire in Callaway Fire Area C-1 0, "Switchgear Room B".

FSAR section 15.5.1.2 describes a bounding design basis accident analysis where key variables are set to conservative values in order to obtain the worst case scenario. The thermal-hydraulic analysis performed for the Fire PRA uses nominal values for plant parameters and the availability of equipment is only limited by the fire damage scenario. The FSAR analysis makes the following conservative assumptions which are not included in the T -H analysis:

1) It uses a "maximum" pump curve in order to conservatively increase the CCP flow rates, whereas the T -H analysis uses a realistic pump curve based on actual performance.
2) In both scenarios the CCP minimum flow recirculation valves are open, therefore the flow into the RCS will be less than the value on the pump curve.
3) Pressurizer level is 43% span, which is 5% higher than the nominal value at the assumed low Tavg of 570.7 degrees F. These assumptions reduce the time needed to fill the Pressurizer.
4) Pressurizer sprays (450 gpm each) are on, which lowers the RCS pressure and consequently increases charging pump flow, thus raising the Pressurizer level. Due to the conservative nature of the spray model used, RCS pressure remains close to the initial value until the time at which filling occurs. This maximizes the injected flow and minimizes the time to fill.
5) Initial Pressurizer pressure is assumed 30 psig lower than the nominal value used in the T-H analysis, which results in higher charging pump flow rates.
6) Initial NSSS power is 102%.
7) The highest worth RCCA is stuck fully withdrawn, resulting in more decay heat which raises the Pressurizer level faster due to swell. The Fire PRA analysis assumes all control rods are inserted.

Page 14 of28 to ULNRC-060 11 The FSAR analysis results show the Pressurizer water inventory increasing by more than 1,000 gpm during the first 8 minutes, therefore the CCP flow rates are only one piece of the overall picture.

ASSUMPTION FSAR 15.5.1.2 C-10 DISCUSSION Safety Injection T=O T=O Same Signal Reactor Trip T=O, assumed with T=O with fire Same SI Two CCP's Pump T=O T=O Same flow initiated Two CCP flow rate approximately 300 130 to 170 gpm FSAR uses stronger gpm based on average (250 gpm pump curve.

maxtmum pump maximum) curve Normal Charging Starts at T=O, NCP tripped by SI FSAR scenario has Pump flow secured at T=6 signal more charging flow.

terminated minutes (secured (Approximate flow manually) rate 50 gpm)

RCS Letdown Isolated at T=O Isolated at T=5 FSAR scenario (letdown flow rate minutes results in smaller 120 gpm) Pressurizer volume to fill.

PORV's Available after 9 Not Available C-1 0 scenario has minutes (by reduced charging operator action) rate due to higher RCS pressure.

Pressurizer Back Up Inoperable (load Unavailable Same Heaters Energize shed with the SIS)

Pressurizer Spray On (450 gpm per Unavailable Reduces RCS valve) pressure which increases CCP flow rates.

Auxiliary Feedwater 2 motor driven and 2 motor driven Slower cool down Pumps one turbine driven feeding 2 S/G's slows charging rate.

feeding all4 S/G's Page 15 of28 to ULNRC-060 11 Given these differences in assumptions and initial conditions, the difference in available time between the FSAR analysis and the T -H analysis is reasonable. The Fire PRA uses the best estimate T -H analysis to develop operator response times.

Page 16 of28 to ULNRC-06011 Section 6: Response to Radiation Release RAis No new RAis were submitted for this topic.

Page 17 of28 to ULNRC-060 11 Section 7: Licensee Identified Changes to the Transition Report LIC-01 Provided by ULNRC-05 851 dated April 17, 2012 LIC-02 Provided by ULNRC-05851 dated April17, 2012 LIC-03 Provided by ULNRC-05 851 dated April 17, 2012 LIC-04 Provided by ULNRC-05851 dated April17, 2012 LIC-05 Provided by ULNRC-05851 dated April17, 2012 LIC-06 Provided by ULNRC-05851 dated April17, 2012 LIC-07 Provided by ULNRC-05851 dated April17, 2012 LIC-08 Provided by ULNRC-05851 dated April17, 2012 LIC-09 Provided by ULNRC-05876 dated July 12,2012 LIC-10 Provided by ULNRC-05876 dated July 12, 2012 LIC-11 Provided by ULNRC-05876 dated July 12, 2012 LIC-12 Provided by ULNRC-05876 dated July 12, 2012 LIC-13 Provided by ULNRC-05876 dated July 12, 2012 LIC-14 Provided by ULNRC-05876 dated Ju!Y_ 12, 2012 LIC-15 Provided by_ ULNRC-05876 dated July 12,2012 LIC-16 Provided by ULNRC-05876 dated July 12, 2012 LIC-17 Provided by ULNRC-05876 dated July 12, 2012 LIC-18 Provided by ULNRC-05876 dated July 12, 2012 LIC-19 Provided by ULNRC-05876 dated July 12,2012 LIC-20 Provided by ULNRC-05952 dated February 19, 2013 LIC-21 The time period for completion of the implementation items in Table S-3 is being changed from 6 months to 8 months. The revised page of LAR Section 5.4 is provided in Attachment 1 of this Enclosure. The revised page of Attachment S is provided in Attachment S of this Enclosure.

Page 18 of28 to ULNRC-06011 Attachment 1: Revisions to the Transition Report Main Body Page 19 of28

Ameren Missouri Callaway Plant NFPA 805 Transition Report criteria set forth in 10 CFR 51 .22(c)(9) for categorical exclusion from the need for an environmental impact assessment or statement.

5.4 Transition Implementation Schedule The following schedule for transitioning Callaway Plant to the new fire protection licensing basis requires NRC approval in accordance with the following schedule:

  • Implementation of new NFPA 805 fire protection program to include procedure changes, process updates, and training of affected plant personnel. This will be the later of 6 months after Refueling Outage 19 (currently scheduled for Spring of 2013) or 8 months I LIC 21 after NRC approval.
  • Attachment S provides a listing of plant modifications associated with the transition to NFPA 805 and their implementation status (open or complete). Currently open modifications will be field completed no later than June 30, 2013. Appropriate compensatory measures for any incomplete NFPA 805 related modifications will be maintained until the modifications are complete.

August 2011 Page 147 to ULNRC-060 11 Attachment A: Revisions to Transition Report Attachment A - NEI 04-02 Table B Transition of Fundamental Fire Protection Program and Design Elements Page 21 of28

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment A. NEI 04-02 Table B-1 -Transition of Fundamental FP Program and Design Elements (NFPA 805 Chapter 3}

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.4.1(b) Industrial fire brigade members shall have Complies, with Required See implementation item identified CAR 201101832, 'Track no other assigned normal plant duties Action below. Implementation Items for NFPA-805 that would prevent immediate response to Project" I All a fire or other emergency as required.

Procedure APA-ZZ-00743, "Fire Team Organization and Duties," Rev. 23/

Section 4.1.3(c)

IMPLEMENTATION ITEMS:

11-805-051 Section 4.1.3(c) of procedure APA-ZZ-00743, "Fire Team Organization and Duties," will be revised to include the requirement that industrial fire brigade members shall have no other assigned normal plant duties that would prevent immediate response to a fire or other emergency as required.

FPE 3.4.1 (c) During every shift, the brigade leader and Complies with The Fire Brigade Leader is a Procedure ODP-ZZ-00001, "Operations at least two brigade members shall have Clarification Licensed Operator or Non- Department- Code of Conduct," Rev.

RAI sufficient training and knowledge of Licensed Operator fully qualified on 82/ Sections 3. 7, 3.9 and 3.11 18 nuclear safety systems to understand the all Operations Technician watch effects of fire and fire suppressants on stations. In addition, two of the four nuclear safety performance criteria. fire brigade members are Non-Exception: Sufficient training and Licensed Operators who, at a knowledge shall be permitted to be minimum, have completed the provided by an operations advisor classroom portion of the primary dedicated to industrial fire brigade and secondary Operations support. Technician watch stations.

3.4.1(d) The industrial fire brigade shall be notified Complies No Additional Clarification Procedure EIP-ZZ-00226, "Fire immediately upon verification of a fire. Response Procedure for Callaway Plant," Rev. 14/ Section 5.2 Procedure OTO-KC-00001 , "Fire Response," Rev. 8/ Step 7 August 2011 PageA-37 to ULNRC-060 11 Attachment H: Revisions to Transition Report Attachment H - NFP A 805 Frequently Asked Question Summary Table Page 23 of28

Ameren Missouri Callaway Plant NFPA 805 Transition Report Table H NEI 04-02 FAQs Utilized in LAR Submittal Closure No. Rev. Title FAQ Ref. Memo 08-0052 0 Transient Fire Growth Rate and ML081500500 ML092120501 Control Room Non-Suppression ML091590505 07-0054* Demonstrating Compliance with ML103510379 ML110140183 Chapter 4 of NFPA 805 09-0056 2 Radioactive Release Transition ML102810600 ML102920405 08-0057 3 New Shutdown Strategy ML100330863 ML100960568 10-0059 5 NFPA 805 Monitoring ML120410589 ML120750108 12-0064 Fire Transient ML121780013 ML122550050 ML12346A488 IPRA RAI 37

  • Note: The FAQ Submittal number was 08-0054 but the NRC Closure Memo for the FAQ was listed as 07-0054. 07-0054 was used to be consistent with the Closure Memo.

August 2011 Page H-3 to ULNRC-060 11 Attachment S: Revisions to Transition Report Attachment S - Plant Modifications and Items to be completed during Implementation Page 25 of28

Ameren Missouri Callaway Plant NFPA 805 Transition Report Table S-3, Items provided below are those items (procedure changes, process updates, and training of affected plant personnel) that will be completed prior to the implementation of the new NFPA 805 FP program. This will be the later of 6 months after Refueling Outage 19 (currently scheduled for Spring of 2013) or 8 months after NRC approval. 1 LIC 21 August 2011 Page S-8

Ameren Missouri Callaway Plant NFPA 805 Transition Report Table S-3 Implementation Items Item Unit Description LAR Section I Source 12-805-004 1 APA-ZZ-00741, "Control of Combustible Materials," will be revised to implement "No Storage" and Attachment V "No Hotwork" controls for Fire Areas A-1, A-11, A-12, A-27, C-1, C-2, C-3, C-7, C-8, C-9, C-10, C-11, C-12, C-17, C-18, C-19, C-20, C-21, C-22, C-23, C-24, C-25, C-26, C-30, C-31, C-32, C-33, C-34, C-36, and C-37.

12-805-005 1 Upon completion of all Fire PRA credited implementation items in Transition Report Table S-2, Attachment W verify the validity of the change-in-risk provided in Attachment W. This includes consideration of the following plant modifications: 05-3029, 07-0151 and 09-0025. If this verification determines that the risk metrics have changed such that the RG 1.205 acceptance guidelines are not met, the new Implementation Item 12-805-005 will require implementation of additional analytical efforts, and/or procedure changes, and/or plant modifications to assure the RG 1.205 risk acceptance criteria are met.

13-805-001 1 Callaway will revise its Fire PRA analysis to utilize alternate analyses specifically by modifying the PRA RAI 36 PRA PRA model to provide detailed HRA of all actions required after Control Room evacuation. This RAI evaluation will be accomplished using the EPRI HRA calculator and be conducted in accordance 36 with the guidance in NUREG-1921 prior to use of the Fire PRA for evaluation of post-transition changes to the Fire Protection Program .

13-805-002 1 Callaway will revise its Fire PRA analysis to utilize the Fractional Influence Factors (e.g., 0.0, 0.1 , PRA RAI 37 PRA 0.3, 1.0, 3.0, 10, or 50) in FAQ 12-0064 and ensure that all evaluated locations have a total RAI influence factor greater than 0.0 prior to use of the Fire PRA for post-transition changes. 37

=----------~------------------~-------------=-------------------------------

13-805-003 Callaway will revise the Fire PRAto utilize the total ignition frequency for Bus Duct fires (BIN 16.2) which are specified in Supplement 1 to NUREG/CR-6850 prior to use of the Fire PRA for PRA RAI 38 IPRA RAI

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. __________________________________________ 38 13-805-004 1 Callaway will revise the Fire PRAto utilize the specific CPT treatment guidance provided in a PRA RAI 39 PRA Letter to Joe G. Giitter from Richard P. Correia dated June 14th, 2013 titled "Interim Technical RAI Guidance on Fire Induced Circuit Failure Mode Likelihood Analysis" prior to use of the Fire PRA 39 for evaluation of post-transition changes to the Fire Protection Program.

13-805-005 Callaway will revise the Fire PRA to utilize the fire growth time to peak heat release rate for trash PRA RAI 40 PRA fires of 8 minutes specified in FAQ 08-0052 incorporated in Supplement 1 to NUREG/CR-6850 RAI (EPRI 1011989) for the Main Control Room transient fires prior to use of the Fire PRA for 40 evaluation of post-transition changes to the Fire Protection Program.

13-805-006 Callaway will implement the uncertainty methodology used in response to PRA RAI 09-B to PRA RAI 41 PRA estimate the change in risk associated with post-transition changes to the Fire Protection Program RAI with the understanding that the uncertainty methodology can be refined to utilize parametric data 41 evaluations.

August 2011 Page S-16

Ameren Missouri Callaway Plant NFPA 805 Transition Report Table S-3 Implementation Items Item Unit Description LAR Section I Source 13-805-007 1 Callaway will revise the Fire PRAto incorporate the Internal Events PRA Common Cause Failures PRARAI42 PRA (CCFs) such that both the current IE and Fire PRA are consistent regarding CCFs prior to using RAI the Fire PRAto evaluate the change in risk associated with post-transition changes to the Fire 42 Protection Program.

August 2011 Page S-17

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