ULNRC-05952, Regarding Adoption of National Fire Protection Association Standard NFPA 805

Regarding Adoption of National Fire Protection Association Standard NFPA 805

ML13051A449
Person / Time
Site:	Callaway
Issue date:	02/19/2013
From:	Maglio S Ameren Missouri, Union Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML130510518	List: ML13051A450 ULNRC-05952, Regarding Adoption of National Fire Protection Association Standard NFPA 805
References
TAC ME7046, ULNRC-05952
Download: ML13051A449 (41)

Text

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WAmeren Callaway Plant MISSOURI February 19, 2013 ULNRC-05952 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 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 of National 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 of National Fire Protection Association Standard NFP A 805 dated July 12, 2012 (TAC No. ME7046)

On March 2, 2012, Union Electric Company (dba Ameren Missouri) formally received from the NRC Staff a request for additional information (RAI) related to the application request to adopt a new fire protection licensing basis based on NFP A 805 which complies with *the requirements in 10 CFR 50.48(a) and (c) and the guidance in Revision 1 of Regulatory Guide (RG) 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants".

On April 17, 2012 Ameren Missouri transmitted to the NRC Staff via Reference 1, responses to a portion of the RAI questions. On April 12th, 17th, & June 21st, 2012, teleconferences between Ameren Missouri and the NRC staff reviewers were conducted in which additional requests/questions on the remaining RAI's were discussed. Following these teleconferences the NRC provided to Ameren Missouri supplemental requests/questions to the original RAI's. On July 12, 2012 Ameren Missouri transmitted to the NRC Staff via Reference 2, responses to the supplemental requests/questions identified during the teleconferences.

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • PO Box 620 Fulton, MD 65251 AmerenMissouri.com

ULNRC-05952 Date February 19,2013 Page2 On December 11, 2012 Ameren Missouri formally received from the NRC Staff a second request for additional information (RAI) related to the application request described above. The purpose of this letter is to provide the responses to the second RAI transmittal dated December 11, 2012 as modified in an email from NRC Staff dated January 22, 2013 (ML13022A024). provides the responses to the second round of RAis. also contains a description of changes to the license amendment request (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.

This submittal does contain new commitments which are provided in the revised LAR Transition Report Attachment S, Plant Modifications and Items to be Completed during Implementation.

The No Significant Hazards Consideration determination provided in the original submittal is not altered by the additional information provided in this letter.

If there are any questions, please contact Mr. Scott Maglio at (573) 676-8719 or Mr. Roger Wink at (314) 225-1561.

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

Sincerely,

~:~

Executed on: YJ_ I 1C1 I 2D l3 Regulatory Affairs Manager GAHa/nls , Request for Additional Information (RAI) with Callaway Plant Response

ULNRC-05952 Date February 19, 2013 Page 3 cc: U.S. Nuclear Regulatory Commission (Original and 1 copy)

Attn: Document Control Desk Washington, DC 20555-0001 Mr. Elmo E. Collins 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-05952 Date February 19, 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 D. W. Neterer L. H. Graessle J. S. Geyer S. A. Maglio R. Holmes-Bobo NSRB Secretary T. B. Elwood Mr. Mike Westman (WCNOC)

Mr. Tim Hope (Luminant Power)

Mr. Ron Barnes (APS)

Mr. Tom Baldwin (PG&E)

Mr. Mike Murray (STPNOC)

Mr. Mark Morgan (SCE)

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

Missouri Public Service Commission Mr. Dru Buntin (DNR) to ULNRC-05952 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: Revisions to Transition Report Attachment E - NEI 04-02 Radioactive Release Transition Attachment F: Not used.

Attachment G: Revisions to Transition Report Attachment G - Recovery Actions Transition Attachment H: Not used.

Attachment I: Revisions to Transition Report Attachment I - Definition of Power Block Attachment J: Not used.

Attachment K: Not used.

Page 1 of37 to ULNRC-05952 Attachment L: Revisions to Transition Report Attachment L - NFP A 805 Chapter 3 Requirements for Approval (10 CFR 50.48(c)(2)(vii))

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 of37 to ULNRC-05952 Section 1: Response to Fire Modeling RAis Fire Modeling RAI 03.01 NFPA 805, Section 2.7.3, "Quality," describes requirements for fire modeling calculations, such as acceptable models, limitations of use, validation of models, defining fire scenarios, etc. This description includes justification of model input parameters, as it is related to limitations of use and validation.

a. At the time of the site audit, additional fire modeling analysis and the reports describing the results of this analysis for Fire Area C-1 ("R1984-001-001 Fire Dynamic Simulator (FDS)

Analysis to Support Detailed Fire Modeling," and "KC-57 Fire Modeling Report for Fire Area C-1") were not officially completed. Since that time, these reports have been completed and reviewed. The latter resulted in the following questions:

1. In Section C 1.4.2 of "R1984-00 1-001, FDS Analysis to Support Detailed Fire Modeling,"

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

n. Section C1.5.7 of"R1984-001-001, FDS Analysis to Support Detailed Fire Modeling,"

states that a design fire of 69 kilowatts (kW) was used in the analysis. Section 7.3.2 of the document entitled, "KC-57 Fire Modeling Report for Fire Area C-1," provides justification for this value. This justification is qualitative and it is not clear how the recommended 98th percentile value of 317 kW was quantitatively scaled to 69 kW. Please describe whether this value was determined based on the method discussed in Section G.5 ofNUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, Final Report,"

September 2005 (ADAMS Accession Nos. ML052580075 and ML052580118, for Volumes 1 and 2, respectively). If so, please describe whether a representative distribution function of this point value of the HRR was developed or was this 'bounding' point value used in conjunction with a severity factor of 1.0 in the Fire PRA. Please provide clarification for the justification that a transient fire in excess of 69 kW cannot occur in this fire zone (Room 31 0 1 in Fire Area C 1).

111. Section C1.5.7 of"R1984-001-001, FDS Analysis to Support Detailed Fire Modeling,"

states that 32 additional pounds of Class A or 13 additional pounds of Class B fire load would be necessary to transition from the maximum expected fire scenario (MEFS) HRR to the limiting fire scenario (LFS) HRR. Please provide additional clarification for how these Page 3 of37 to ULNRC-05952 estimates were made. Please explain what administrative controls are, or will be, in place to ensure these relatively small additional fire loads will not be present in this particular space.

In addition, provide justification for the HRR of the MEFS used in the analysis.

tv. For the HRRs used in the analysis, a t2 profile was used for a transient fire that reached its peak at 8 minutes, as recommended in FAQ-08-0052. The time to peak HRR used in the analysis is recommended for common trash cans such as plastic or metal receptacles. The suggested time to peak HRR for trash fires that are not contained in plastic or receptacles is 2 minutes, which is more representative of a "no storage" area trash fire. Additionally, the suggested time to peak HRR for an oil spill (Class B) fire is 0 minutes. Please explain what administrative controls are, or will be, put in place to prevent loose trash and liquid fuel spill fires in this area.

b. Note that the wording for Fire Modeling RAI 03.01 b was revised following a clarification phone call between the NRC staff and Ameren Missouri on 1/15/13. The revised RAI text as provided by the NRC to Ameren Missouri in an email dated 1122/13 (ADAMS Accession No. ML13022A024) is as follows:

Based on the staffs independent calculations, it appears that the plume radius for scenario C-31.3618-8 has been overestimated. This is likely to be the case for all other scenarios. Confirm that the use of the plume radii estimates do not have an adverse effect on the results of the risk calculations.

c. The detailed fire modeling reports of several fire areas (e.g., A 11, C 1, C21, and C31) refer to the MEFS and the LFS. The terms MEFS and LFS are typically used when fire modeling is performed to support performance-based evaluations in accordance with NFPA 805, Section 4.2.4.1. However, Section 4.5.1.2 in the LAR states that "Fire modeling was performed as part of the FPRA development (NFPA 805 Section 4.2.4.2)." Please provide confirmation that this statement in the LAR is correct and that no fire modeling was performed to support compliance with NFPA 80S. Section 4.2.4.1.

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

Response to Fire Modeling RAI 03.01 a.i. The normalized parameter for radial distance relative to fire diameter is not applicable in this fire scenario because the target HDPE pipe is outside of the zone of influence (ZOI) of the radiative heat flux. The lowest point of the HDPE pipe is 12 feet above floor level. The transient fire is postulated to occur 2 feet above floor level. The modeled fire will produce a peak flame height of less than 3 feet. The critical heat flux for the HDPE pipe is not available from the manufacturer; however, using a critical heat flux of 3 kw/m2 (the critical heat flux Page 4 of37 to ULNRC-05952 for solid state components, significantly less conservative than that for HDPE piping), the heat flux ZOI has a maximum radius of less than 3 feet. Based on these calculations, and at 12 feet above floor level, the HDPE pipe is well above the ZOI of the fire.

Since the normalized parameter related to radial distance relative to fire diameter is relevant only when determining the sensitivity of fire modeling results (in this case, target damage due to radiative heat flux) to the input parameters (distance from the fire), and damage due to radiative heat flux is not possible, this normalized parameter is not applicable.

a.ii. The transient heat release rate was determined based on the guidance in NUREG/CR-6850 Section G.5 and the decisions of the EPRI-led Fire PRA Methods Review Panel. The methodology is detailed in the Ameren Missouri response to PRA RAI-23 as submitted to the NRC in letter ULNRC-05876 dated July 12,2012.

The 69kW heat release rate was modeled as a bounding point value fire size using a severity factor of 1.0.

The analysis of Fire Area C-1, Fire Zone 31 01, using the methods referenced above, determined that a bounding 69kW transient heat release rate was justified based on the following:

• Fire Zone 3101 will be subject to strict combustible controls (designated as "No Storage")

and so paper, cardboard, scrap wood, rags and other trash will not be allowed to accumulate in the area. The implementation of a "No Storage" area designation is being tracked via Item 12-805-004 in Transition Report Table S-3 "Implementation Items". The revised Transition Report Table S-3 is provided in AttachmentS of this enclosure.

• Large combustible liquid fires are not expected in this fire zone since activities in the areas do not include maintenance of oil containing equipment.

• The transient fire history in the plant was reviewed and a transient fire has not occurred in this fire zone.

• A transient fire in an area of strict combustible controls, where only small amounts of contained trash are considered possible, is judged to be no larger than the 75th percentile fire in an electrical cabinet with one bundle of qualified cable.

• The materials composing the fuel packages included in Table G-7 ofNUREG/CR-6850 (e.g., eucalyptus duff, one quart of acetone, 5.9kg of methyl alcohol, etc.) are not representative of the typical materials expected to be located in these areas.

• A review of the transient ignition source tests in Table G-7 ofNUREG/CR-6850 indicates that of the type of transient fires that can be expected in these rooms (i.e., polyethylene trash can or bucket containing rags and paper) were measured at peak heat release rates of 50kW or below.

Page 5 of37 to ULNRC-05952 Since only small quantities of trash in temporary containers can be expected, a 69k W peak heat release rate was determined to be appropriate to represent this quantity of combustibles.

The 69kW heat release rate bounds the small trash can fires reported in NUREG/CR-6850 Appendix G.

a.iii. The calculation of necessary material to transition from MEFS to LFS consisted of determining the required fuel mass based on the fire size and duration (e.g. 69kW over 60 minutes) in BTUs. The BTU value was divided by the heat of combustion for Class A and Class B materials (in BTU/lb) to obtain the mass of the Class A or Class B material required to sustain the indicated fire size and duration.

Class A materials typically consist of wood, paper, and cable insulation. The heat of combustion for these materials was estimated to be 8,000 BTU/lb., based on materials of this type listed in Table 3-4.16 ofthe Society of Fire Protection Engineering (SFPE) Handbook, 4th Edition. Using this value, the required additional fuel load to transition from MEFS (69kW) to LFS (142kW) was calculated to be approximately 32 pounds.

Class B materials typically consist of plastics and oil. The heat of combustion for these materials was estimated to be 20,000 BTU/lb., based on materials of this type listed in Table 3-4.16 of the SFPE Handbook, 4th Edition. Using this value, the required additional fuel load to transition from MEFS (69kW) to LFS (142kW) was calculated to be approximately 13 pounds.

Plant procedure APA-ZZ-00741, "Control of Combustible Materials," will be revised to implement "No Storage" and "No Hot Work" locations credited in the Fire PRA, including Fire Area C-1. This revision is being tracked by Implementation Item 12-805-004 and has been added to AttachmentS, Table S-3, to the Transition Report. The controls will ensure the additional fire loads are prevented in the area.

The MEFS used in the analysis was a 69kW HRR fire. The basis for use of a 69kW fire in this area is discussed in the response to FM RAI 03.01a-ii.

a.iv. Plant procedure APA-ZZ-00741, "Control of Combustible Materials," will be revised to implement "No Storage" and "No Hot Work" locations credited in the Fire PRA, including Fire Area C-1. This revision is being tracked by Implementation Item 12-805-004 and has been added to Transition Report AttachmentS Table S-3. The revised page is included in Attachment S to this enclosure.

b. The plume radius, as calculated by Callaway Plant, is conservative. The use and validation of the plume radius correlation was provided in the Ameren Missouri response to Fire Modeling RAI 01.d as submitted to the NRC in letter ULNRC-05876 dated July 12, 2012. Since the plume radius was applied in a conservative manner, there is no adverse effect on the results and conclusions of the risk calculations.

Page 6 of37 to ULNRC-05952

c. The statement in the Callaway Plant License Amendment Request (LAR) Transition Report Section 4.5.1.2 "Fire modeling was performed as part of the FPRA development (NFPA 805, Section 4.2.4.2)" is correct. No fire modeling was performed to support compliance with NFPA 805, Section 4.2.4.1.

The maximum expected fire scenario (MEFS) and limiting fire scenario (LFS) are used to assist in performing an assessment of safety margins in fire modeling. The MEFS determines, by fire modeling, whether performance criteria are met in the fire compartment being analyzed. The LFSs are those that result in unfavorable consequences with respect to the performance criteria being considered. In essence, the output for the LFS calculations is obtained by manipulating the fire model input parameters until consequences are obtained that violate the target damage limits established. Thus, the LFS can be based on a maximum possible, though very unlikely, value for one input variable, or an unlikely combination of input variables. The goal of determining an LFS is to be able to analyze the margin between these scenarios and those used to establish the MEFS.

The actual evaluation of the margin between the MEFS and the LFS can be largely qualitative, but it provides a means of identifying weaknesses in the analysis where a small change in a model input could indicate an unacceptable change in the consequences.

In general, a qualitative safety margin review is performed. Based on the implementation of a conservative approach and the assumptions outlined in the safety margin review, sufficient margin has been built into the fire modeling calculations up front. Therefore, limiting fire scenarios have not been generated for most scenarios.

Page 7 of37

Enclosure 1 to ULNRC-05952 Section 2: Response to Fire Protection Engineering RAis Fire Protection Engineering RAI 15.01 In a letter dated April 17, 2012, the licensee responded to Fire Protection Engineering RAI 15 regarding high hose station pressures, and discussed the previous removal of pressure reducers as a result of Unresolved Item (URI) #483/87018-03. Based on this response, the referenced URI, and Calculation KC-27, additional information is required:

a. Typically, a fire hose is service tested to 150 pounds per square inch (psi). The letter referenced above indicates operating pressures up to 160 psi. Please clarify that the service testing pressure for fire hoses adequately accounts for this higher pressure (e.g., service testing exceeds the maximum operating pressure - usually 11 0 percent or 50 psi over the maximum operating pressure).

b. The closure of URI #483/870 18-03 required the posting of signs on hose stations with high pressures. Please clarify the continued adequacy of such signs for operating pressures up to 160 psi.

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

Response to Fire Protection Engineering RAI 15.01

a. Callaway Plant procedure MPM-KC-FW004 18 Month Fire Hose Station Inspection Outside Areas, is used to perform service testing of hose rack fire hose. The service test pressure used is that which is stamped on the hose which is 250 psi.

b. A walk down of the Callaway Plant hose stations was performed prior to responsding to RAI 15 and it was verified the signage warning of high system pressures required by the URI are in place on hose stations.

c. The letter dated April17, 2012 (ULNRC-05851) contains the correct reference which is URI

1. 483/87018-03. The typographical error in KC-27 will be corrected in the next routine update however due to the nature of the change no commitment is deemed necessary.

Fire Protection Engineering RAI 16 LAR Table I-1 : Power Block Definition lists power block structures within the Owner Controlled Area that were determined to contain equipment required to meet the nuclear safety and radioactive release performance criteria. In comparing those structures listed in Frequently Asked Question (FAQ) 06-0019 (ADAMS Accession No. ML080510224), and NEI 04-02, "Guidance for Implementing a Risk-Informed, Performance Based Fire Protection program Under 10 CFR50.48(c)," Section K, the NRC staff noticed certain structures were not listed in LAR Table I-1.

Page 8 of37 to ULNRC-05952 Please ensure those structures listed within NEI 04-02 Attachment K and FAQ-06-00 19 are accounted for as either within or not within the power block (e.g., Service Building, Water Treatment, Intake Structure, etc.). In addition, please clarify that the Hot Machine Shop, if one exists, and the YARD should also be within the power block. For the YARD, please identify the specific structures or equipment to be within the power block, which may include fire pump house, transformers, YARD equipment managed by plant staff, warehouses, and other significant structures containing equipment required for operations. Please revise LAR Table I-1 as necessary.

Response to Fire Protection Engineering RAI 16 In response to Fire Protection Engineering RAI 16, the Callaway Plant Transition Report Attachment I has been revised to specifically address the following: the Service Building, Water Treatment, Intake Structures, the Hot Machine Shop, and YARD structures. The changes to Transition Report Attachment I are shown in Attachment I to this enclosure.

Fire Protection Engineering RAI 17 NFPA 805 Section 3.3.5.1 requires cables located above suspended ceilings to be listed for plenum use, routed in armored cable, routed in metallic conduit, or routed in cable trays. LAR Attachment L, Approval Request 2, has identified certain areas and cables that do not meet NFP A 805 Section 3.3.5.1 and provided justifications to deviate from such requirements.

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

b. Please clarify if there are any fire detection systems installed above suspended ceilings in these areas of concern. For those cables that do not meet the NFPA 805 Section 3.3 .5 .1 criteria, please describe the separation between the power, control, and lighting cables and the telephone/data/communication cables, if any.

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

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

e. The regulations in 10 CFR 50.48(c)(2)(vii) require that radioactive release performance measures must be addressed by the request as well as nuclear safety performance measures.

Please revise the request to address the radioactive release performance measures (goals, objectives, and performance criteria). Please include in your discussion how the non-fire Page 9 of37 to ULNRC-05952 water use could potentially impact firefighting efforts and how this might impact liquid and gaseous releases.

f. The Safety Margin section states that "the safety margin inherent in the analysis for the fire event has been preserved." Please clarify that this performance-based method does not change the assumptions and limitations of the analytical methods used in the development of the FPP.

Response to Fire Protection Engineering RAI 17 In response to the RAI, the LAR Transition Report Attachment L, Approval Request 2 has been revised as follows:

• Enhanced the description of the plant areas applicable to the request to address RAI 17 parts a and b.

• The typographical error noted in part c has been corrected.

• LAR Transition Report Attachment L, Approval Request 2 text and LAR Transition Report Attachment A Table B-1 Section 3.3.5.1 have been revised to address RAI 17 part d and Implementation Item 11-805-050 revised to require future compliance with NFPA 805 Section 3.3.5.1.

• RAI parts e and fare addressed by changes to the discussion on Radioactive Release, the Basis for Request section and Safety Margin, as shown in Attachment L of this enclosure.

Page 10 of37

Enclosure 1 to ULNRC-05952 Section 3: Response to Programmatic RAis Programmatic RAI 01 Section 2.7.3, "Quality," ofNFPA 805 has the following specific requirements:

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

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

2. 7.3 .3: Acceptable engineering methods and numerical models shall only be used for applications to the extent these methods have been subject to verification and validation. These engineering methods shall only be applied within the scope, limitations, and assumptions prescribed for that method.

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

2. 7.3 .5: An uncertainty analysis shall be performed to provide reasonable assurance that the performance criteria have been met."

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

Response to Programmatic RAI 01 The Callaway Plant Transition Report Section 4. 7.3 has been revised to address future analyses and provide specific compliance statements. The revised pages of Transition Report Section 4. 7.3 are provided in Attachment 1 to this enclosure.

Page 11 of37 to ULNRC-05952 Section 4: Response to Safe Shutdown RAis Safe Shutdown RAI 07 Databases and software that integrate fire protection program structure, system, and component data; fire modeling results, and PRA analyses (e.g., EPM-SAFE-PB and ARC) have a range of uses applicable to NFP A 805 implementation. These uses are subject to several NFP A 805 requirements including those that address determination of success paths; completion of the Nuclear Safety Capability Assessment (NSCA); the quality, configuration control, documentation, and verification and validation of analyses; and limitations of use. In addition, these databases and software can be used to facilitate integration of several aspects of NFPA 805 compliance. Specific applicable NFP A 805 requirements include:

NFP A 805 Section 2.2.9 "Plant Change Evaluation" states that: "In the event of a change to a previously approved fire protection program element, a risk informed plant change evaluation shall be performed and the results used as described in 2.4.4 to ensure that the public risk associated with fire-induced nuclear fuel damage accidents is low and that adequate defense-in-depth and safety margins are maintained."

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

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

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

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

NFPA 805 Content requirements include:

(2. 7 .1.1) "The analyses performed to demonstrate compliance with this standard shall be documented for each nuclear power plant (NPP). The intent of the documentation is that the assumptions be Page 12 of37 to ULNRC-05952 clearly defined and that the results be easily understood, that results be clearly and consistently described, and that sufficient detail be provided to allow future review of the entire analyses.

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

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

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

NFP A 805 configuration control requirements include:

(2. 7.2.1) "The design basis document shall be maintained up-to-date as a controlled document.

Changes affecting the design, operation, or maintenance of the plant shall be reviewed to determine if these changes impact the fire protection program documentation."

(2. 7 .2.2) "Detailed supporting information shall be retrievable records. Records shall be revised as needed to maintain the principal documentation up-to-date."

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

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

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

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

(2.7.3.4) "Cognizant personnel who use and apply engineering analysis and numerical models (e.g.,

fire modeling techniques) shall be competent in that field and experienced in the application of these methods as they relate to nuclear power plants, nuclear power plant fire protection, and power plant operations." "An uncertainty analysis shall be performed to provide reasonable assurance that the performance criteria have been met."

The NRC staff notes that, given the broad range of requirements applicable to use of integration databases and software, the Transition Report provided insufficient details for the staff to complete its review of the various areas affected by this software. Therefore, the staff requests that the following additional information be provided, Page 13 of37 to ULNRC-05952

a. A description of how the post transition change evaluation process will ensure that the potential interfaces between integration databases and software and other databases and analyses (e.g., the cable and raceway database, the NSCA, the fire PRA, and fire modeling) are evaluated and updated, as appropriate.

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

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

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

Response to Safe Shutdown RAI 07

a. Changes to the plant will be screened for potential NFPA-805 impact using the Engineering Screen/Hazards Review form STARS-ENG-5001-8.3 and the Engineering Screen/Programs Review form STARS-ENG-5001-8.4. These forms contain low level screening questions that have the Engineer contact the Subject Matter Experts (Fire Protection Engineer, SAFE Engineer, or PRA Engineer) as needed for further evaluation. The Engineer qualified to modify the SAFE software will decide if the proposed change requires an update to the Fire Safety Analyses, SAFE software or NSCA analysis, including CARTS (Cable And Raceway Tracking System) information changes in Director. The Fire Protection Engineer will evaluate proposed changes for any impact on the Fire Safety Analyses, fire modeling calculations, or credited Recovery Actions that are part of the NSCA analysis or Fire PRA and initiate an update. The Fire PRA Engineer will evaluate proposed changes for potential impact on the Fire Risk Evaluations, Fire Safety Analyses or the Fire PRA and make any needed updates.

b. The post transition change process summarized in the response to question a. above will ensure that integration databases and software are maintained with up to date plant configuration data. Maintenance and updates to the PRA models are governed by Callaway Plant procedure APA-ZZ-00312, Probabilistic Risk Assessment, which requires that the models are maintained with up to date plant configuration data.

c. Calculations are performed in accordance with procedure EDP-ZZ-04023, Calculations, which requires an independent review. Engineers who perform calculations at Callaway Plant must be qualified to Nuclear Engineering Qualification Standard T62. 7304, Perform and Review Calculations. The integration databases and software are controlled by procedure EDP-ZZ-040 11, Nuclear Engineering Analytical Software Controls, which includes verification and validation requirements. Engineers who perform detailed fire modeling at Callaway Plant must be qualified to Qualification Standard T62. 7118Q, Perform Detailed Fire Page 14 of37 to ULNRC-05952 Modeling, which ensures that detailed fire modeling is performed by persons who have the proper training and experience. Implementation Item 11-805-72 will develop a Fire PRA qualification standard to ensure that Fire PRA activities are performed by properly trained and experienced personnel. A qualification standard will also be developed for Engineers who will update the SAFE database.

d. Calculations are performed in accordance with procedure EDP-ZZ-04023, Calculations, which requires an independent review. The integration databases and software are controlled by procedure EDP-ZZ-040 11, Nuclear Engineering Analytical Software Controls, which includes verification and validation requirements. Detailed fire modeling is performed in accordance with Engineering Design Guide ME-0 14, Detailed Fire Modeling. PRA calculations are governed by procedure AP A-ZZ-00312, Probabilistic Risk Assessment (PRA), which specifies independent review and quality requirements.

Page 15 of37

Enclosure 1 to ULNRC-05952 Section 5: Response to Probabilistic Risk Assessment (PRA)

RAis PRARAI24 Please describe how the evaluation includes the possible increase in heat release rate (HRR) caused by the spread of a fire from the ignition source to other combustibles. Please summarize how suppression is included in the evaluation.

Response to PRA RAI 24 The increase in heat release rate (HRR) caused by the spread of a fire from the ignition source to other combustibles was calculated by combining each individual fire HRR as a function of time, and then using the combined total HRR as the input to the zone-of-influence models. Heat release rates from NUREG/CR-6850 were used, and the rules for propagation to cable trays and fire spread rates follow the guidance of the FLASH-CAT model found in NUREG/CR-7010. This approach is described in more detail in the Ameren Missouri response to Fire Modeling RAI 03a, as submitted to the NRC in letter ULNRC-05876 dated July 12, 2012.

Suppression is credited in the Detailed Fire Modeling, Multi-Compartment and Main Control Room Fire Analyses.

For each ignition source, the fire point of origin was determined by field inspection. Distances from the point of origin to the nearest fixed suppression/detection device and to the ceiling were used as inputs to the NUREG-1805 FDTlO spreadsheets, to calculate the device activation time. The HRR input in the spreadsheet was varied to determine the minimum HRR required to activate the suppression/detection device. The fire growth and propagation analysis was then used to determine the time to reach the minimum HRR (required for suppression/detection). This minimum HRR includes the ignition source, propagation to adjacent cabinet vertical sections and spread/propagation to secondary cable trays, as appropriate.

For some areas, suppression activation was determined using Fire Dynamics Simulator (FDS). The justification for the use of FDS to determine suppression activation is described in the Ameren Missouri response to Fire Modeling RAI 03e, as submitted to the NRC in letter ULNRC-05876 dated July 12, 2012.

The Detailed Fire Modeling analysis credited suppression to arrest fire growth and propagation, thereby limiting target damage. Once automatic suppression activates or manual suppression is initiated, further fire growth and propagation is not postulated. The Zone of Influence (ZOI) at the time to suppression was calculated and used to determine target damage. The time to suppression includes time to activate devices, any necessary delays for pre-discharge alarms, and an agent discharge time. Suppression timing and target damage sets for suppression success and failure are documented in the fire area specific Detailed Fire Modeling reports.

Page 16 of37

Enclosure 1 to ULNRC-05952 Suppression was credited in the Multi-Compartment Analysis (MCA) to mitigate the spread of fire from the initiating area to adjacent fire areas. Suppression was credited for all areas and scenarios with the potential to produce a hot gas layer (HGL). In some cases, suppression was credited based on the results of the detailed fire model. However, the MCA also credits suppression in areas where no detailed fire modeling was performed but a HGL was deemed possible due to hazards in the room.

The methodology and results ofthis analysis are documented in Calculation 17671-010c, "Callaway NFPA 805 Fire PRA- Multi-Compartment Analysis," Revision 1.

Suppression was credited in the Main Control Room Fire Analysis for main control board and electrical cabinet fire scenarios. The analysis determined the time to fire suppression for use as an input to the Control Room evacuation probability. The analysis is documented in Calculation 17671-010b, "Callaway NFPA 805 Fire PRA- Main Control Room Fire Analysis," Revision 1.

In all cases, non-suppression probabilities were developed according to the guidance ofNUREG/CR-6850 and FAQ 08-0050.

PRARAI25 Transient fires should at a minimum be placed in locations within the plant physical access units (PADs) where conditional core damage probabilities (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 cabling associated with each. Transient fires should be placed at all appropriate locations in a PAU where they can threaten pinch points. Hot work should be assumed to occur in locations where hot work is a possibility, even if improbable (but not impossible), keeping in mind the same philosophy. Please describe how transient and hot-work fires are distributed within the PAUs. In particular, please identify the criterion that determines where an ignition source is placed within the PAUs. Also, if there are areas within a PAU where no transient or hot-work fires are located since those areas are considered inaccessible, please define the criteria used to define "inaccessible." Note that an inaccessible area is not the same as a location where fire is simply unlikely, even if highly improbable.

Response to PRA RAJ 25 Transient fires have been postulated in each fire compartment in the Fire PRA. All available floor area is postulated as a possible transient ignition source location. Some compartments have been sub-divided into transient zones (weighted by floor area), to refine the frequency of damage to risk-significant targets. The total transient and hot work ignition frequency is apportioned throughout the available floor area.

In some cases, a fraction of the floor area may have been considered inaccessible in a fire compartment when visual inspection confirmed that either:

• The floor area is obstructed due to the presence of equipment (e.g., floor area on which an electrical panel is located), or Page 17 of37

Enclosure 1 to ULNRC-05952

• Personnel are unable to access the area due to structural obstructions (e.g., columns/walls) or equipment location (e.g., floor based piping configurations prevent access to the floor area).

Locked or high radiation areas are not considered inaccessible. The Fire PRA determined that the only area in which transient fires are precluded during power operation is Fire Zone RB 1 in Fire Area RB-I. This fire zone represents the RCP cubicles and the space within the reactor building secondary shield wall, all of which are inaccessible while at-power. This has been documented in Calculation 17671-005 "Callaway NFPA 805 Fire PRA- Ignition Frequencies," Revision 1.

PRARAI26 The transition report describes and justifies an initial coping time of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, after which, actions are necessary to maintain safe and stable beyond 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. Please provide a discussion of the actions necessary during and beyond 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> to maintain safe and stable conditions beyond 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> such as refilling fluid tanks or re-aligning systems.

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

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

Response to PRA RAI 26 The PRA models are quantified for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The PRA definition of a successful endstate requires that the core cooling is being provided at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and can continue to be provided in the near term.

Success criteria for the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> includes all relevant water inventories, AC power and DC power. It includes automatic and/ or manual actions to provide alternate water sources if necessary. The Callaway Plant Transition Report discussion in Section 4.2.1.2 regarding the results of the Safe and Stable evaluation has been revised to include a discussion of the FPRA treatment of Safe and Stable.

The revised pages of Transition Report Section 4.2.1.2 are provided in Attachment 1 to this enclosure.

PRARAI27 Please describe how fire-induced instrument failure (including no readings, off-scale readings, and incorrect/misleading readings) is addressed in the fire human reliability analysis (HRA) for new operator actions (not including post MCR abandonment which are addressed elsewhere) that have been credited in the risk estimates.

Page 18 of37 to ULNRC-05952 Response to PRA RAJ 27 The fire human reliability analysis considered fire-induced instrumentation failure (including no readings, off-scale readings, and incorrect/misleading readings) at two points in the Fire PRA.

The first point was in the evaluation of the undesired response to spurious instrumentation failures as part of ASME/ANS-RA-Sa-2009 supporting requirement ES-C2. The second point was in the quantification of human error probabilities modeling the operator response to a fire-induced initiating event. The treatment of fire-induced instrumentation failure for both these cases is summarized below.

1. For cases with no instrumentation readings due to fire impact on the instrumentation (including loss of power supplies to the instruments), any operator action directly requiring the failed instrumentation for diagnosis was not credited in the fire PRA, i.e., the HEP was set to a failure probability of 1.0.

2. For cases where instrumentation required for an operator action is degraded (for example, one instrumentation train failed and one instrumentation train unaffected), the HEP is increased to reflect the additional diagnosis effort that may be required using the guidance of NUREG-1921. Typically, this is applied to systems with redundant instrumentation channels where all redundant channels are not showing the same value, requiring additional interpretation from the operators.

3. For off-scale/incorrect/misleading ("spurious") readings the concerns are: (1) that operators will not take an action that is required or (2) operators will take actions that are not required and that could aggravate the response.

a. For the first concern where operators need to take an action that relies on instrumentation, and the required instrumentation is failed by the fire, the fire HRA quantifies the human error probability as 1.0 (guaranteed failure) as described in Item 1 above. If the operators need to take an action that relies on instrumentation that is degraded by the fire (such as one channel is off-scale, incorrect, or misleading and one channel is protected), then the fire HRA quantifies the HEP as higher than the case with all instrumentation available, as described in Item 2 above.

b. The second concern was treated in the evaluation of supporting requirement ES-C2 (from Chapter 4 of ASME/ANS-RA-Sa-2009). In the evaluation, potential cognitive errors of commission were identified by a systematic review of the operating and alarm procedures within the context of the PRA accident scenarios. As a first step in this identification process, the alarm procedures were reviewed to identify any that required immediate responses without verification of the alarm condition, e.g., a low lubrication oil pressure alarm for a turbine driven pump requires operators to immediately trip the pump. If the cabling associated with this alarm is impacted by fire in a given scenario or the cable was not traced, it was assumed that the alarm would annunciate and prompt the operators to follow the alarm procedure. All alarm procedures were screened out from potentially causing an undesired operator response because each procedure required confirmation before taking the action. After the alarm procedure review, the second step in the evaluation was to review the emergency operating procedures to identify any instructions Page 19 of37 to ULNRC-05952 that may lead to inappropriate actions given spurious instrumentation failure. Such actions were then systematically screened based on diversity of instrumentation or being inconsequential. At Callaway Plant, all such actions have been screened.

PRARAI28 Please identify if any variance from deterministic requirements (VFDRs) in the license amendment request (LAR) involved performance-based evaluations of wrapped or embedded cables. If applicable, please describe how wrapped or embedded cables were modeled in the fire PRA (FPRA) including assumptions and insights on how the PRA modeling of these cables contributes to the VFDR delta-risk evaluations.

Response to PRA RAJ 28 Cables protected by Electrical Raceway Fire Barrier Systems (ERFBS) were credited in the FPRA as being protected from fire damage, commensurate with the fire barrier rating of the ERFBS. If the ERFBS were determined to have a fire rating that exceeded the fire duration of a fire scenario, the cable was considered protected free of fire damage and did not contribute to risk evaluations. In total there were two instances where a VFDR was identified for ERFBS, as follows:

• VFDR A-06-001: In Fire Area A-6, raceways 1J1L01 and 1U1K01 are provided with a Darmatt fire wrap with an installation (i.e., framing) that results in the fire rating of the ERFBS to be degraded from the intended 3-hour rating of the design criteria to a 1-hour rating. This condition represents a variance from the deterministic requirements ofNFP A 805, Section 4.2.3 and is a degraded barrier issue. The detailed fire modeling determined that the raceways are not impacted by fire events in that area. The 1-hour rating of the fire wrap is not credited in the FPRA and has been credited for defense-in-depth purposes only. The analysis is summarized in Calculation KC-86 "Fire Safety Analysis for Fire Area A-6," and has determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.

• VFDR A-18-0001: In Fire Area A-18, the Darmatt 1-hour rated fire wrap for conduit 111097 is degraded (i.e., notched). The fire rating ofthis ERFBS is degraded from the intended 1-hour rating of the design criteria. This condition represents a variance from the deterministic requirements ofNFPA 805, Section 4.2.3 and is a degraded barrier issue. The degraded wrap was not credited to provide any fire resistance rating. The detailed fire modeling determined that the raceway was subject to damage in a transient fire scenario. The delta risk calculations include failure of this raceway in the applicable scenarios. The analysis is summarized in Calculation KC-98 "Fire Safety Analysis for Fire Area A-18," and has determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria ofNFPA 805 Section 4.2.4.

In addition, for each ERFBS protected cable, a review was performed to determine compliance with Supporting Requirement No. FSS-C8 of ASME-ANS RA-Sa-2009, which requires that all credited ERFBS shall not be subject to either mechanical damage or direct flame impingement from a high-hazard ignition source. Compliance with Supporting Requirement No. FSS-C8 of ASME-ANS RA-Page 20 of37 to ULNRC-05952 Sa-2009 was established and is documented in the compartment specific Detailed Fire Modeling reports.

There are no VFDRs involving performance based evaluations of embedded cables. Raceways and cable embedded in concrete were credited in the FPRA as being protected from fire damage, commensurate with the fire barrier rating of the embedment. The rating of concrete coverage is documented in Callaway Plant Calculation KC-26,"Nuclear Safety Capability Assessment".

Embedded raceway and cables that were determined to be protected with an embedment fire rating that exceeded the fire duration of a fire scenario did not contribute to risk evaluations.

PRARAI29 Please identify any changes made to the internal events PRA (lEPRA) or FPRA since the last full-scope peer review of each of these PRA models that are consistent with the definition of a "PRA upgrade" in American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS)-

RA-Sa-2009, "Standard for Levell/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," as endorsed by NRC Regulatory Guide (RG) 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities." Also, please address the following:

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

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

Response to PRA RAI 29 Internal Events PRA (lEPRA)

The Callaway Plant lEPRA underwent a full-scope review against the Capability Category II requirements of ASME RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," with ASME RA-Sa-2003 and ASME RA-Sb-2005 Addenda, ASME, 2005, in 2006. The results of that review, and the impact of those results on the Callaway Plant NFPA 805 Fire PRA, are summarized in Attachment U of the Transition Report.

A major revision to the lEPRA was made, in part, to address F&Os from the 2006 review against the Standard. Following completion of this revision, changes to the lEPRA that were deemed to constitute a "PRA upgrade" (consistent with the definition in the Standard) were reviewed during a focused-scope peer review against Capability Category II requirements of relevant Supporting Requirements of ASME/ANS RA-Sa-2009, including any Clarification and Qualifications provided in Regulatory Guide 1.200, Revision 2. The focused-scope peer review was conducted in October 2011 and led by Page 21 of37 to ULNRC-05952 Westinghouse Electric Corporation. The following items were the upgraded pieces of the lEPRA that were reviewed in the focused-scope peer review:

• Support system initiating event fault tree models

• Interfacing systems LOCA modeling

• Incorporation of the WOG 2000 RCP seal LOCA model

• Implementation of an expanded (e.g., MGL) common cause failure methodology

• Revised LERF model

• Revised internal flooding analysis Findings from the focused-scope peer review and their resolution are addressed in the PRA Quality Report (17671-015).

Additionally, in May 2011 a focused scope peer review was conducted in order to verify a PRA upgrade regarding HRA methods. This focused scope review did not produce any Findings and the results are summarized in Attachment 5 of the PRA Quality Report (17671-015).

Fire PRA CFPRA)

A full-scope Fire PRA Peer Review was conducted in November 2009. The results of this peer review are discussed in Attachment V of the Transition Report. Since November 2009 no changes were made to the Fire PRA which would be classified as "Upgrades". Model changes were made to the FPRA to address the F&O's and model changes were made to comply with FAQ's which were finalized after Nov 2009.

PRARAI30 Please identify any plant modifications (implementation items) in Attachment S of the LAR that have not been completed but which have been credited directly or indirectly in the change-in-risk estimates provided in Attachment W. When the effects of a plant modification have been included in the PRA before the modification has been completed, the models and values used in the PRA are necessarily estimates based on current plans. The as-built facility after the modification is completed may be different than the plans. Please add an implementation item that, upon completion of all PRA credited implementation items, verifies the validity of the reported change-in-risk. This item should include a plan of action should the as-built change-in-risk exceed the estimates reported in the LAR.

Response to PRA RAI 30 Attachment S of the Transition Report identifies plant modifications and items to be completed during implementation. Tables S-1 and S-2 identify plant modifications completed and committed to, Page 22 of37 to ULNRC-05952 respectively. Both of these tables include a column indicating whether the modification is currently included in the FPRA. Thus, those modifications with a "Y" in the "In FPRA" column of Table S-2 represent modifications that have not been completed, but have been directly credited in the FPRA.

Additionally, those modifications with an "N" in the "In FPRA" column of Table S-2 represent modifications that have not been completed, but potentially impact risk as they may have an indirect impact on the FPRA. Table S-2 was reviewed and the specific modification items that may impact the risk metrics of Attachment W of the transition report are Modifications 05-3029, 07-0151 and 09-0025.

Implementation Item 12-805-005 has been added to Table S-3 to verify the validity of the reported change-in-risk estimates of Attachment W following completion of these modifications. The revised S-3 Table is included in AttachmentS to this enclosure. 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.

PRARAI31 Please describe whether changes to the seismicities as a result of the United States Geological Survey (USGS) re-evaluation for the central and eastern United States (USGS, "2008 NSHM Gridded Data, Peak Ground Acceleration"), based on reanalysis of the New Madrid earthquakes, were considered in determining the applicability of the seismic-fire interaction analysis performed for the individual plant examination for external events (IPEEE) to the current state of seismic-fire interactions. If not, please describe how the changes would affect the conclusions made regarding seismic-fire interactions.

Response to PRA RAJ 31 The USGS seismicity parameter changes do not affect seismic-fire interaction conclusions in either the Callaway Plant IPEEE analysis or the current (NFPA 805 Fire PRA Task 12) analysis, as both analyses were performed qualitatively.

The current seismic-fire interaction analysis uses and relies on the IPEEE analysis. The conclusion section of the Callaway Plant IPEEE seismic/fire assessment states, in part, the following:

"From the pre-screening results and walkdowns performed, Callaway appears to be a very seismic resistant plant. In general, most structures and equipment were overqualified for seismic events ...

During the pre-screening it was evident that vendors tended to overtest most items for loads that were over specified such that very large seismic margin exists. During the walkdown it was observed that most equipment was well anchored and supported so large margins exist relative to the RLE. "

Based on the above conclusion, and the qualitative nature of both the current and IPEEE seismic-fire interaction analyses, USGS seismic hazard changes, as compared to the IPEEE Review Level Page 23 of37 to ULNRC-05952 Earthquake (RLE), would not affect the conclusions of the Callaway Plant NFP A 805 Fire PRA seismic-fire interaction analysis.

PRARAI 32 Please discuss the calculation of the frequencies of transient and hot-work fires. Please characterize the use of the influence factors for maintenance, occupancy, and storage, noting if the rating "3" is the most common, as it is intended to be representative of the "typical" weight for each influence factor.

It is expected that the influence factor for each location bin associated with transient or hot-work fires will utilize a range of influence factors about the rating "3," including the maximum 10 (or 50 for maintenance) and, if appropriate, even the rating "0." Note that no PAU may have a combined weight of zero unless it is physically inaccessible, administrative controls notwithstanding. In assigning influence factor ratings, those factors for the Control/Auxiliary/Reactor Building are distinct from the Turbine Building; thus, the influence factor ratings for each location bin are to be viewed according to the bin itself.

Response to PRA RAI 32 The Callaway Plant Fire PRA distributed transient fires over 376 rooms in the plant, using the NUREG/CR-6850 methods and data with 2 previously documented exceptions (PRA RAJ 8b). The rooms were grouped into 4 plant areas as indicated in the table below:

Plant Area as defined in Number of Rooms at NUREG/CR-6850 Callaw~y in this Area Containment 5 Plant Wide Components 130 Control/Auxiliary Building 175 Turbine Building 66 TOTAL 376 The distribution scheme used the NUREG/CR-6850 weighting factors with the exception that a weighting factor of0.05 was used for Maintenance in rooms with a hotwork prohibition and a weighting factor of 0.1 was used for Storage in those rooms designated as "transient combustible free zones". Occupancy did not use any value less than 1.0. The table below shows the weighting factor distributions over the plant-wide total of 376 rooms.

Page 24 of37 to ULNRC-05952 TOTAL PLANT WEIGHTING FACTORS OVER 376 ROOMS

1. Rooms Assigned this #Rooms Assigned #Rooms Assigned Weighting factor for this factor for this factor for Factor Maintenance Occupancy Storage 0 1 1 1 0.05 29 0 0 0.1 0 0 29 1 191 215 173 3 152 147 134 10 3 13 39 Average Weighting 1.8 2.1 2.6 Factor The next 4 tables show the weighting factor distribution in each of the four plant areas as indicated in NUREG/CR-6850.

CONTAINMENT WEIGHTING FACTORS (5 ROOMS)

1. Rooms Assigned this #Rooms Assigned #Rooms Assigned Weighting factor for this factor for this factor for Factor Maintenance Occupancy Storage 0 1 1 1 1 4 4 4 Average Weighting 0.8 0.8 0.8 Factor PLANT-WIDE WEIGHTING FACTORS (130 ROOMS)

1. Rooms Assigned this #Rooms Assigned #Rooms Assigned Weighting factor for this factor for this factor for Factor Maintenance Occupancy Storage 1 76 89 76 3 52 41 49 10 2 0 5 Average Weighting 1.9 1.6 2.1 Factor Page 25 of37 to ULNRC-05952 CONTROL BUILDING WEIGHTING FACTORS (175 ROOMS)

1. Rooms Assigned this #Rooms Assigned #Rooms Assigned Weighting factor for this factor for this factor for Factor Maintenance Occupancy Storage 0.05 29 0 0 0.1 0 0 29 1 77 97 68 3 68 65 46 10 1 13 32 Average Weighting 1.7 2.4 3.0 Factor TURBINE BUILDING WEIGHTING FACTORS (66 ROOMS)

1. Rooms Assigned this #Rooms Assigned #Rooms Assigned Weighting factor for this factor for this factor for Factor Maintenance Occupancy Storage 1 34 25 25 3 32 41 39 10 0 0 2 Average Weighting 2.0 2.2 2.5 Factor Discussion of RAI Queries:

The specific questions in the RAI are discussed below. The overall consideration is that this represents a weighting scheme to distribute transient frequency to some rooms relative to others. Total transient bin frequency from NUREG/CR-6850 is preserved regardless of what weighting scheme is used. Different weighting factors will distribute transient frequency differently to the rooms, but the total bin frequency is preserved. The specific responses to items requested in this RAI are provided below:

A. THE 50 FACTOR: The 50 weighting factor was not used for any Callaway Plant fire area or room. This causes the frequency distribution to be narrower than it would be if a 50 was used.

If a 50 had been used, there is no indication that it would be in a high risk room (i.e., cable chases or cable spreading rooms), but rather it would be in the I&C hot shop (A-26) or the Page 26 of37 to ULNRC-05952 diesel generator rooms (D-1, D-2). If a 50 had been used, it would have most likely re-distributed risk from a high-risk room to a lower risk room.

B. THE MOST COMMON FACTOR: On a plant wide basis, the most frequently used weighting factor is "1 ",closely followed by a weighting factor of"3". In the turbine building, the "3" weighting factor is more common than the "1" weighting factor. It is not clear from NUREG/CR-6850 if it matters which weighting factor is most often assigned.

C. THE AVERAGE FACTOR: The average weighting factor is less than 3, except for the Control Building Storage. The average weighting factor for Callaway Plant is between 1. 7 and 3.0 (except for the Containment building). It is not clear in NUREG/CR-6850 if"3" is intended to be the "average" weighting factor or the most commonly used weighting factor.

Because the weighting factors progress in a geometric progression rather than an arithmetic progression, the average weighting factor will not be "3" unless the number of rooms attributed to each weighting factor is predetermined to make the average "3 ". If all the weighting factors have an equal number of rooms assigned, the average will be 4.67 (if "50" is not used). If the weighting factors have rooms assigned in a symmetrical distribution, the average weighting factor is also 4.67. In order to have an average weighting factor of"3", the number of rooms assigned to each weighting factor would have to be in inverse proportion to the progression of the weighting factors themselves. This type of effort was not done for Callaway Plant.

D. SPECIAL FACTORS: A special weighting factor of0.05 was used for Maintenance in hotwork prohibited zones and a 0.1 was used for Storage in transient combustible free zones.

These weighting factors are not specified in NUREG/CR-6850, but it was necessary to have a factor less than 1.0 to account for the administrative controls to reduce transient and welding fires. Except for one room in Containment, the minimum value for occupancy was 1.0, thus the combined weighting factors were always greater than 1.0.

E. ROOMS WITH A "0.0": One room in Containment was assigned a 0.0 for all three factors.

This is the reactor coolant pump (RCP) room. 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. This value is inconsistent with the ASME standard, but the impact on risk is negligible. The RCP area has a fixed fire frequency of 2.35E-3/yr. due to the reactor coolant pumps. Adding a transient frequency of 5.8E-4/yr. (i.e., normal occupancy, storage and maintenance) would not be expected to have any impact on risk.

F. CONTAINMENT BUILDING: The four non-zero areas in the Containment building are all assigned the same weighting factors, which are 1-1-1. The result is that all areas are assigned one fourth of the bin 3 frequency for transient fires. Because all areas were assessed to have the same transient characteristics, the assignment of an actual weighting factor is immaterial.

All areas are judged to have the same transient frequency.

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Enclosure I to ULNRC-05952 PRARAI33 By letter dated July 12, 2012, the licensee responded to PRA RAI Ol(c) regarding the effect of updated common cause failure (CCF) probabilities from the recent lEPRA update, and stated that, despite the presence of increases in some CCF probabilities, "a succinct sensitivity study for each CCF value to determine its individual effect was not possible within the time constraints of this RAI response .. . The updated CCFs are scheduled to be incorporated as part of the next fire PRA revision." While this fact and observation (F&O) (1-13) is "only" a suggestion with respect to the lEPRA from the recent focused-scope peer review, it remains unknown as to the potential effect on the conclusions made for transition based on the FPRA. Please provide, at a minimum, a bounding analysis to estimate the potential effect upon fire core damage frequency (CDF), large early release frequency (LERF), delta(~) CDF and~ LERF, as this is necessary to confirm that the conclusions remain the same.

Response to PRA RAI 33 A sensitivity analysis has been conducted to estimate the change in total CDF and delta-CDF calculated via the FPRA using updated common cause failure probabilities from the lEPRA. The sensitivity was only performed on CDF. CDF is the limiting risk metric at Callaway Plant, because LERF is always lower than 10% ofCDF. If the CDF has acceptable change, it can be inferred that LERF also has acceptable change. The sensitivity analysis was conducted in two parts as follows.

The first part was to evaluate the CDF increase for those basic events already modeled in the Fire PRA. This step in the sensitivity analysis consisted of a spreadsheet calculation in which the CCF events in the FPRA and lEPRA were matched, and then the differences in the probabilities (lEPRA versus FPRA) were multiplied by the FPRA Birnbaum importances to produce the change in CDF and the change in delta-CDF associated with each CCF event. The individual CCF basic event changes to CDF were then summed to determine an overall FPRA change in CDF. Similarly, the impact of individual CCF basic event changes was summed to calculate the change in delta-CDF. For all the CCF events that have a direct match between the lEPRA and the Fire PRA, the net change in both CDF and delta CDF was negative. Total CDF decreased 1.38E-6 per year and delta CDF decreased 9.19E-8 per year.

The second part was to evaluate the CDF increase for those basic events that are not modeled in the Fire PRA. The second part was conducted in two steps. The first step was to calculate the change in variant risk and the second was to relate the change in variant risk to the change in delta-risk, as a bounding risk approach. Each of these steps is described below.

There is one set ofCCF events in the current lEPRA which are not included in the FPRA. These 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 was accomplished by using surrogate events (NSAFP test and maintenance events). The impact of the CCF of the safety-related and non-safety AFW pumps was bounded by assuming the non-safety auxiliary feedwater pump is failed. If a basic event for the non-safety pump is set to 1.0, then the basic events for CCF of the safety-related and non-safety AFW pumps would be inconsequential, as they are non-minimal.

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Enclosure 1 to ULNRC-05952 The risk achievement worth (RAW) of the non-safety AFW pump test and maintenance term in the total CDF equation is 1.6, representing an increase of 1.24E-5 per year for the fire CDF resulting in a total fire CDF of3.24E-5 per year. Similarly, the risk achievement worth (RAW) of the non-safety AFW pump test and maintenance (T&M) term in the variant CDF equation is 2.50. This results in an increase in the variant fire CDF of9.22E-6 per year, for a total variant fire CDF of 1.54E-5 per year.

Note that, 1.54E-5 is not "delta risk as defined by RG 1.205", but is "variant risk" as defined below.

The variant risk equation is the total risk equation for all scenarios which include a VFDR. The variant risk equation represents the bounding delta-risk equation as it assumes that all compliant risk is zero. When using Boolean cutset equations, it is impractical to write a delta-risk equation other than the bounding case. The delta risk (as defined by RG 1.205) is the difference between the "variant risk equation with all credited recovery actions" and the "variant risk equation with VFDR cutsets removed". For Callaway Plant the variant CDF is 6.15E-6 per year, while the "delta risk" as calculated in the FRE's is 2.03E-6. The variant CDF as calculated by WinNUPRA is approximately 3 times as high as the delta-CDF value as defined in RG 1.205. During the development of the transition report, the delta CDF was calculated for each scenario during the Fire Risk Evaluation process. The total delta CDF in the transition report is 2.03E-6 per year. For Callaway Plant, the ratio of the delta risk to the variant risk is 2.03E-6/6.145E-6 = 0.33. Therefore, if the variant risk increases to 1.54E-5 per year as described above, the delta risk would increase to 5.09E-6 per year. The 5.09E-6 per year is the estimate of the bounding risk increase, and this is still under the RG 1.205 acceptance value of 1E-5 per year. LERF risk metrics were not included in the sensitivity study, since the Callaway Plant FPRA-determined risk is dominated by core damage risk.

Based upon the above-described sensitivity analyses, the new (i.e., lEPRA) CCF probabilities will not change the conclusions made for transition based on the existing FPRA.

PRARAI34 By letter dated April 17, 2012, the licensee responded to PRA RAI 02, and stated that "The Fire PRA Peer Review is therefore consistent with the clarifications in RG 1.200, Rev. 2." If correct, please add the following" ... the clarifications AND QUALIFICATIONS in ... " If not, please describe and justify why it is not consistent with "qualifications."

Response to PRA RAI 34 Omission of "and qualifications" from the cited sentence in the April 17, 2012 letter was simply an oversight. Therefore, the FPRA portion of the response to PRA RAI 02 in the April 17, 2012 letter is hereby revised to the following:

The Fire PRA Peer Review team (October 2009) used the clarifications and qualifications to the PRA standard as presented in Regulatory Guide (RG) 1.200, Revision 2. The Westinghouse Owner's Group (WOG) Fire Peer Review training is required for every peer reviewer prior to participating in an industry Fire PRA peer review. The training documents specifically instruct peer reviewers to consider the clarifications and qualifications ofRG 1.200 during the review process. Additionally, the database used for the peer review process during the Westinghouse Peer Reviews includes the Page 29 of37 to ULNRC-05952 most up to date RG 1.200 clarifications and qualifications, which facilitates and emphasizes their inclusion during the review. The Fire PRA Peer Review is therefore consistent with the clarifications and qualifications in RG 1.200, Rev 2.

PRARAI35 The NRC has revised the wording for PRA RAJ 35 per NRC Staff email dated January 22, 2013:

The April17, 2012, response to PRA RAJ 07b (ADAMS Accession No. ML12108A239, ML12108A240),justifies the use of a CCDP ofO.l and CLERP of0.01 for alternate shutdown where this failure probability represents both failures of equipment and operator actions. The justification for these CCDP and CLERP values is based on a qualitative feasibility assessment of the operator actions, which consists of a qualitative argument that the actions have been determined to be feasible.

It may be acceptable to take the position that operator actions are dominant in the CCDP and CLERP.

However, no quantitative assessment of CCDP and CLERP was provided to verify the CCDP of 0.1 and CLERP of 0.01 given that operator actions dominate. Despite feasibility considerations being addressed, it is not obvious that a CCDP value ofO.l (and CLERP = 0.01) represents the failure probability of an action of this complexity. Provide further justification for the 0.1 and 0.01 by providing the results of the human failure event (HFE) quantification process described in Section 5 ofNUREG-1921, considering the following

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

2. The results ofthe process in Section 5.2. 7 ofNUREG-1921 for assigning scoping human error probabilities (HEPs) to actions associated with switchover of control to an alternate shutdown location. The bases for the answers to each of the questions asked in Figure 5-4 should be addressed.

3. The results of the process in Sections 5.2. 8 of NUREG-1921 for assigning scoping human error probabilities (HEPs) to actions for performing alternate shutdown once switchover is complete. The bases for the answers to each of the questions asked in Figure 5.5 should be addressed.

4. The results of a detailed HRA quantification, per Section 5.3 of NUREG-1921 in place of items 2 and 3 if a CCDP as low as 0.1 (and CLERP as low as 0.01) is not attainable through the scoping approach. For the detailed study, quantify the contribution via the evaluation of different scenarios upon MCR evacuation, including the sum of those scenarios in the results for the CCDP and CLERP.

Provide a sensitivity analysis that shows the impact on the PRA results (CDF, LERF, ~CDF, ~LERF) of using the resultant CCDP/CLERP analysis for control room abandonment scenarios.

Response to PRA RAJ 35 The response to PRA-RAJ-07b needs to be clarified. PRA-RAJ-07b stated that a probability of 0.1 was used for the "total probability of failing to evacuate and establish local control successfully". The value of 0.1 was used as an overall Human Error Probability for all human error events necessary to Page 30 of37 to ULNRC-05952 provide safe shutdown outside the control room. In calculating the conditional core damage probability (CCDP), failure of equipment was accounted for independently of the HEP. The total CCDP is 1.2E-1. For the conditional Large Early Release (LER) probability (CLERP), the highest LERF split fraction was appended to each cutset in the CCDP equation. The CLERP is 3.37E-3.

Rather than develop a detailed HRA analysis of a highly uncertain issue, this RAI response will investigate the CDF/LERF margin for successively higher HEP values. Therefore, to address this RAI, sensitivity analyses have been performed as described below.

A. The first sensitivity case examines the impact of a bounding human error probability (HEP).

The MCR Abandonment HEP was set to failed (probability of 1.0), and thus a conditional core damage probability (CCDP) of 1.0 was assigned to all control room evacuation scenarios.

Using this HEP/CCDP, the CDF for the main control room increases from 7.8E-7 per year to 1.38E-6 per year.

B. A more-realistic HEP/CCDP of 0.5 was applied to all control room evacuation scenarios. In this case, the main control room CDF increased from 7.8E-7 per year to 1.04E-6 per year.

For each of these sensitivity cases, Callaway Plant would continue to meet the Regulatory Guide 1.205 risk acceptance criteria.

One of the reasons the previous Callaway Plant response referenced the plant's feasibility evaluation was to provide sufficient qualitative justification that the action was credible. The feasibility evaluation addresses potential failure modes related to performance shaping factor issues such timing, procedures, training, staffing, and communications. There are limitations with existing HRA methods to quantify the MCR Abandonment HEP and capture the complexities of the plant response, so Callaway Plant has chosen to treat this issue as a two case, bounding sensitivity. The plant's model of record uses a MCR Abandonment HEP of0.1, considers 0.5 to be an upper bound, and has quantified with an extreme bound of 1.0. In all cases, Callaway Plant meets the Regulatory Guide 1.205 risk acceptance criteria for transition.

Note that the favorable results of these two sensitivities are due primarily to the low evacuation probabilities for panel fires in the control room, and the train separation of the cable spreading rooms.

Also note that the sensitivities use the revised (July, 2012) CFAST analysis of the MCR, which incorporates all previous RAI concerns.

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Enclosure 1 to ULNRC-05952 Section 6: Response to Radiation Release RAis Radiation Release RAI 01 For liquid releases, Column 4 of Table E-2 indicates the Fire Areas where liquid effluents are collected in monitored tanks. For each area, Column 4 also states that Fire Pre-Plan steps provide reasonable assurance liquids are contained within the fire area. For each such area, please provide a qualitative assessment describing the following:

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

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

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

d. The duration of anticipated fire fighting activities;

e. The anticipated amount of water to be generated;

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

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

Response to Radiation Release RAJ 01 As discussed during a January 15th, 2013 telephone call with the NRC staff, and consistent with Frequently Asked Question (FAQ) 09-0056, this response provides a description of the overall capabilities for directing, monitoring, and processing potentially contaminated liquid effluent from firefighting activities in lieu of a discussion of the above attributes for each specific area.

Floor Drains In the monitored areas, the floor drains do not have a direct route to the outside. The effluent from those drains is processed and monitored in the Radwaste Building prior to discharge. Water flows initially into two Floor Drain Tanks each with a capacity of 10,000 gallons. Should these tanks become full they can be discharged into two larger Discharge Monitor Tanks which have a capacity of 100,000 gallons each. This overall capacity provides reasonable assurance that the floor drain system will hold firefighting hose stream runoff and/or fixed suppression system discharge for fires within monitored locations. In the subject plant areas with fixed suppression the floor drains have been sized to accommodate the fixed suppression system runoff.

Page 32 of37 to ULNRC-05952 Fire Brigade Actions In accordance with Implementation Item 11-805-076, the following action will be taken; Applicable fire preplans will be updated to accomplish the following:

• Modify the fire brigades primary and secondary access entry points to limit opening doors to non-monitored areas.

• Add a statement in the Precautions section that is read by the Incident Commander that all water discharges must be monitored prior to discharge to a non-monitored location.

In accordance with Implementation items 11-805-077, 078, 079, 081, 082, 083 and 084, applicable Fire Brigade lesson plans and training will be updated to add discussions of radiological considerations in the handling and discharge of fire suppression related liquid runoff.

These steps will ensure the fire brigade is knowledgeable and aware of the need to contain contaminated fire suppression runoff associated with fire scenarios within monitored areas and the fire pre-plans provide guidance to reinforce that training and knowledge.

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

Response to Radiation Release RAI 02 In accordance with Implementation Item 11-805-076, the following action will be taken; Applicable fire preplans will be updated to accomplish the following:

• Modify the fire brigade's primary and secondary access entry points to limit opening doors to non-monitored areas.

• Add a statement in the Precautions section that is read by the Incident Commander that all smoke and water discharges must be monitored prior to discharge to a non-monitored location.

• Add a statement in the Ventilation section that all smoke must be monitored for contamination prior to exhaust to a non-monitored location (outdoors).

In accordance with Implementation items 11-805-077, 078, 079, 081, 082, 083 and 084, applicable Fire Brigade lesson plans and training will be updated to add discussions of radiological considerations in the fire ventilation decision process.

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Enclosure I to ULNRC-05952 These steps will ensure the fire brigade is knowledgeable and aware of the need to contain contaminated gaseous products within monitored areas and the fire pre-plans provide guidance to reinforce that training and knowledge.

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

Response to Radiation Release RAI 03 Fire areas A-21 and A-22 are physically located within the Auxiliary Building at elevation 2047' and they contain the Main Control Room HVAC related equipment. For ventilation purposes only they are part of the Main Control Room HVAC system envelope.

Per Implementation Item 11-805-076, the Fire Preplan Manual will be updated for fire areas A-21 and A-22 to include precautions for the Fire Brigade and Control Room personnel:

• Ventilation in this fire area is part of the Control Room ventilation. A fire in this area may cause smoke to be discharged into the Control Room. Upon detection of a fire in this area immediately secure affected train Control Room ventilation.

• Monitor smoke and water prior to discharge to areas outside the Auxiliary Building.

The Fire Preplan Manual will also contain the following instructions in the Ventilation section for both fire areas:

• Contact RP (Radiation Protection) to monitor smoke for contamination before ventilating outside the Auxiliary Building. If contamination is detected use Auxiliary/Fuel Building normal exhaust to remove smoke. If contamination is not detected use portable fans and ducting to ventilate out the NORTH missile door into the Turbine Building.

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

Page 34 of37 to ULNRC-05952 Response to Radiation Release RAJ 04 Fire Areas C-5 through C-8 are located within the Control Building 1984' elevation and, contrary to the statements in Table E-2, these fire areas are provided with a separate stand-alone HVAC system which is filtered, monitored and exhausts via the unit vent. Table E-2 has been corrected to address the error and reference added to the Access Control Exhaust system which is a filtered and monitored pathway. The revised pages of Transition Report Table E-2 are provided in Attachment E to this enclosure.

Implementation Item 11-805-076 will revise the Fire Preplan Manual for these fire areas to contain the following instructions in the Ventilation section for the fire areas in question:

• Contact RP to monitor smoke for contamination before ventilating outside the Control Building. If contamination is detected use Access Control Exhaust to remove the smoke. If contamination is not detected use portable fans and ducting to ventilate up the stairwell to the outside.

Radiation Release RAI 05 For Areas TB-1 on page E-31 , and YD-1 on page E-32, Columns 4 (Liquid Effluents) and 5 (Gaseous Effluents) of Table E-2 reference Callaway Plant Calculation KC-43, "NFPA 805 Code Comparison."

Section 4.4 for evaluations performed for specific locations within the Non-Radiologically Controlled Areas (RCAs) ofTB-1 and YD-1. For these two areas, a bounding analysis, Callaway Plant Calculation HPCI 10-04, is also referenced in Column 7 (Conclusions) of the table. Column 7 indicates that that the bounding analysis and administrative controls along with Fire Pre-Plan steps limits radiation release due to direct effects of fire suppression activities. It also states that in addition to compliance with NFP A 805 Sections 4.2.3 and 4.4.4, it satisfies the performance requirements of NFPA 805 for radiation release. Please provide copies of the referenced documents, administrative controls and Fire Pre-plan steps and provide justification as to how the release will not only not exceed the limits in NFPA 805 but also not exceed the instantaneous gaseous effluent dose rate limits and the liquid effluent concentration limits in the plant's Technical Specifications (TS) (Reference FAQ 0056, ADAMS Accession No. ML102920405.)

Response to Radiation Release RAJ 05 As discussed during a January 15th, 2013 telephone call with the NRC staff, in lieu of submitting the requested documents Callaway Plant Calculation KC-43, "NFPA 805 Code Comparison" and Callaway Plant Calculation HPCI 10-04, "National Fire Protection Association (NFPA) Standard 805 Airborne and Liquid Effluents Offsite Dose", a description of the documents in the context of the question will be provided. Additionally, the documents are available for review on the Callaway Plant information portal.

Per FAQ 09-0056 Closure Memo (ML102920405), compliance with the radioactive release goals, objectives, and performance criteria can be demonstrated by review of engineering controls to ensure containment of gaseous and liquid effluents. Otherwise, provide an analysis that demonstrates the Page 35 of37 to ULNRC-05952 limitations for instantaneous release of radioactive effluents specific in each unit's Technical Specifications are met. The NRC staff position is that the limitations in a licensee's Technical Specifications are structured to maintain the 10 CFR Part 20 limits and meet the NFP A 805 Radioactive Release Performance Criteria.

As noted, the Callaway Plant fire areas YD-1 (Yard) and portions of TB-1 (Turbine Building) are areas that do not rely on engineering controls for containment of gaseous and liquid effluents. For these fire areas the responding Incident Commander and fire brigade will assess the fire scenario to determine if there is a potential for gaseous and liquid radiological effluent. If that potential exists, in accordance with training and fire preplan guidance, support from Radiation Protection personnel for monitoring and containment action will be requested. Callaway Plant Implementation Items 11-805-076, 080,081, 082, and 083, document the update of fire brigade training and fire pre-plans.

Immediate actions would be based on the specific fire scenario and controls for liquid and gaseous effluents will be utilized to the extent possible. The spill would then be addressed by the actions specified in HDP-ZZ-07000, Radiological Monitoring Program and Groundwater Protection Initiative, which specifies additional containment, monitoring, and reporting actions beyond those taken during the immediate fire event.

To provide an additional level of assurance, a bounding worst case offsite dose analysis in the form of calculation HPCI 10-04, "National Fire Protection Association (NFPA) Standard 805 Airborne and Liquid Effluents Offsite Dose" was developed to evaluate radioactive material storage (RAM) and the effects of fire and water surcharge and to establish maximum allowed RAM storage dose limits. The activity limits specified provide reasonable assurance any assumed fire could not result in exceeding the performance criteria ofNFPA 805 and 10CFR20 limits.

For gaseous effluents the calculation establishes maximum dose limits for RAM storage using the source term of an Intermodal container, a B-25 Box, and a 55 gal drum whose contents are consumed by a fire from a worst case location. The total quantity of radioactivity in the involved container was assumed to be instantaneously released as airborne radioactivity and transported across the site boundary. Administrative controls to adhere to the calculation limits for each container type are embedded within the RAM storage program and survey processes. The dose to the Member of the Public from airborne radionuclides released to the receptor location is less than the Technical Specification limit of 7.5 mrem to any organ. The dose to the Member of the Public was determined in accordance with the methodology and parameters in the Callaway Plant Offsite Dose Calculation Manual (ODCM). Based on the dose limitations placed on RAM storage, gaseous effluents will not exceed Callaway Plant Radiological Effluent Controls (REC) Section 16.11.2.3.a which bounds the limit of 100 mrem promulgated in 10CFR20.

The calculation also reviewed the postulated effects of liquid effluents. It was assumed that contaminated liquids (water, foam, etc.) used in the firefighting effort are not contained and are allowed to reach native soil. The source term developed for the assessment of airborne releases was used to assess the dose from a release of liquid discharged during the firefighting activity. The analysis concluded that based on site specific hydrological conditions and liquid runoff paths, liquid effluent from fire-fighting activities in locations outside the permanent RCA will not exceed Callaway Plant REC Section 16.11.1.2.a which bounds the limit of 100 mrem promulgated in 10CFR20.

Page 36 of37 to ULNRC-05952 Section 7: Licensee Identified Changes to the Transition Report LIC-01 Provided by_ ULNRC-05851 dated April17, 2012 LIC-02 Provided by ULNRC-05851 dated April17, 2012 LIC-03 Provided by ULNRC-05851 dated April17, 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 July 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 Transition Report Attachment C Table B-3 was previously revised to include VFDR RB-02-002 per NRC Staff question #51. Table G-1 should have also been updated, but was overlooked. This licensee identified change will update Table G-1 to include VFDR RB-02-002 in Area RB-1 for BGHV8149A. The revised page of Transition Report Table G-1 is provided in Attachment G of this enclosure.

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