ML12194A638

Response to Request for Additional Information Re Adoption of National Fire Protection Association Standard NFPA 805

ML12194A638
Person / Time
Site:	Callaway
Issue date:	07/12/2012
From:	Ameren Missouri, Union Electric Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML121940471	List: ML12194A638
References
ULNRC-05876
Download: ML12194A638 (237)

Text

Enclosure 1 to ULNRC-05876 Enclosure 1, Request for Additional Information (RAI) with Callaway Plant Response Section 1: Response to Fire Modeling RAIs Section 2: Response to Fire Protection RAIs Section 3: Response to Monitoring Program RAIs Section 4: Response to Safe Shutdown RAIs Section 5: Response to Probabilistic Risk Assessment RAIs Section 6: 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: Revisions to Transition Report Attachment B - NEI 04-02 Table B Nuclear Safety Capability Assessment - Methodology Review Attachment C: Revisions to Transition Report Attachment C - NEI 04-02 Table B Fire Area Transition Attachment D: Revisions to Transition Report Attachment D - Non-Power Operational Modes Transition Attachment E: Not used.

Attachment F: Not used.

Attachment G: Not used Attachment H: Revisions to Transition Report Attachment H - NFPA 805 Frequently Asked Question Summary Table Page 1 of 136 to ULNRC-05876 Attachment I: Not used.

Attachment J: Revisions to Transition Report Attachment J - Fire Modeling V&V Attachment K: Not used.

Attachment L: Revisions to Transition Report Attachment L - NFPA 805 Chapter 3 Requirements for Approval (10 CFR 50.48(c)(2)(vii))

Attachment M: Not used.

Attachment N: Not used.

Attachment O: Not used.

Attachment P: Not used.

Attachment Q: Not used.

Attachment R: Not used.

Attachment S: Revisions to Transition Report Attachment S - Plant Modifications and Items to be completed during Implementation Attachment T: Revisions to Transition Report Attachment T - Clarification of Prior NRC Approvals Attachment U: Not used.

Attachment V: Revisions to Transition Report Attachment V - Fire PRA Quality Attachment W: Revisions to Transition Report Attachment W - Fire PRA Insights Attachment X: Revisions to Transition Report Attachment X - Other Requests for Approval Page 2 of 136 to ULNRC-05876 Section 1: Response to Fire Modeling RAIs Fire Modeling RAI 01 Section 2.7.3.2, "Verification and Validation," of NFPA 805 states: "Each calculational model or numerical method used shall be verified and validated through comparison to test results or comparison to other acceptable models."

Section 4.5.1.2 of the Transition Report of the LAR states that a fire modeling study was performed as part of the fire probabilistic risk assessment (FPRA) development (NFPA 805, Section 4.2.4.2).

During the audit, the NRC staff noted that the fire modeling that was done in support of the LAR was in the form of a plant-specific Fire Modeling Database (FMDB), called, Transient Analysis Worksheets." The FMDB was developed in lieu of using NUREG-1805, "Fire Dynamics Tools (FDTs): Quantitative Fire Hazard Analysis Methods for the U.S. Nuclear Regulatory Commission Fire Protection Inspection Program," December 2004 (ADAMS Accession No. ML043290075)

(FDTs) or NUREG-1824, "Verification &Validation of Selected Fire Models for Nuclear Power Plant Applications, Volume 4: Fire-Induced Vulnerability Evaluation (FIVE-Rev1)," May 2007 (FIVE-Rev1) (ADAMS Accession No. ML071730499) (FIVE-Rev1).

Regarding the verification and validation of the fire models:

a. Please describe how FMDB -Transient Analysis Worksheets were verified (i.e., how was it ensured that the empirical equations/correlations were coded correctly and that the solutions are identical to those that would be obtained with the corresponding chapters in NUREG-1805 or FIVE-Rev1?).

b. The fire models that were used in support of the FPRA are listed in Section 4.5.1.2 of the Transition Report and reference is made to Attachment J of the Transition Report for a discussion of the acceptability of the listed fire models. For the following models, Attachment J states, in part, that "V&V was documented in NUREG-1824," and that "the correlation is used within the limits of its range of applicability."

• Flame Height (Method of Heskestad)

• Plume Centerline Temperature (Method of Heskestad)

• Radiant Heat Flux (Point Source Method)

• Hot Gas Layer (Method of MQH)

• Hot Gas Layer (Method of Beyler)

• Hot Gas Layer (Method of Foote, Pagni, and Alvares [FPA])

• Hot Gas Layer (Method of Deal and Beyler)

• Ceiling Jet Temperature (Method of Alpert)

• Smoke Detection Actuation Correlation (Method of Heskestad and Delichatsios)

The fact that a correlation is used within its range of applicability does not guarantee that it is applied within the validated range reported in NUREG-1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications," May 2007 (ADAMS Accession Page 3 of 136 to ULNRC-05876 No. ML071650546). Please provide technical details to demonstrate that the correlation has been applied within the validated range or to justify the application of the correlation outside the validated range reported in NUREG-1824.

Additional Clarification Needed To demonstrate that the models are applied within the range of applicability, the normalized parameters described in NUREG 1824 and 1934 were calculated and the calculations showed that the parameter is within the validated range or justification was provided for using a parameter value outside the validated range.

• Fire Froude Number: Explain in more detail how you determined quantitatively, the effect of using a conservative convection fraction of 0.7 on the safety margin in the plume ZOI calculation. In addition, describe the criteria that were used to judge that the margin of safety is sufficient.

• Ceiling Jet Radial Distance Relative to Ceiling Height: Provide the normalized parameter values and discuss in light of the sprinkler activation calculations.

• Compartment Aspect Ratio: Confirm that the correct damage temperature in each of the fire zones where the sensitivity analysis was conducted is in fact 330 °C and not 205

°C.

For the case of fires which postulate propagation to secondary combustibles and a Froude number above the validation range, the licensee indicates the fire conditions are calculated assuming a nominal base area which is smaller than the area of the ignition source. Provide examples where this assumption was made.

c. Attachment J of the Transition Report states that the following models are verified and validated on the basis that they are described in an authoritative publication in fire protection literature:

• Heat Detection Actuation Correlation

• Sprinkler Activation Correlation

• Corner and Wall Heat Release Rate

• Correlation for Heat Release Rates of Cables (Method of Lee)

• Correlation for Flame Spread over Horizontal Cable Trays (FLASH-CAT)

Furthermore, the Transition Report states that these models are used within their range of applicability, which does not guarantee that they are applied within the validated range. Please provide technical details to demonstrate that the model has been applied within its validated range or to justify the application of the model outside its validated range.

Additional Clarification Needed To demonstrate that the models are applied within the range of applicability the normalized parameters described in NUREG 1824 and 1934 were calculated and either showed that the Page 4 of 136 to ULNRC-05876 parameter is within the validated range or justification was provided for using a parameter value outside the validated range.

• Sprinkler Activation Correlation: Explain in more detail how the Ceiling Jet Distance Ratio normalized parameter was calculated. The draft response describes the r/H calculation used in the sensitivity analysis. Is the H equal to the distance between the floor and ceiling or the distance between the ignition source and the ceiling? In addition, for Fire Area A-16, explain in more detail how the critical heat release rate for sprinkler activation was calculated in FDT 10.

Indicate exactly where in NUREG 1824 it is stated that the total heat release rate was used in the validation of Alperts ceiling jet correlation.

d. Attachment J of the Transition Report states that the "Plume Radius (Method of Heskestad) model is verified and validated on the basis that it is described in an authoritative publication in the fire protection literature. Please provide technical details to demonstrate that the model has been applied within its validated range or to justify the application of the model outside its validated range.

e. Attachment J of the Transition Report states that the verification and validation of the following applications of Fire Dynamics Simulator (FDS) are documented in NUREG-1824.

• Hot Gas Layer (HGL) Calculations using FDS

• Sprinkler Actuation Calculation using FDS

• Temperature Sensitive Equipment Zone of Influence Study using FDS

• Plume/Hot Gas Layer Interaction Study using FDS Please provide technical documentation that demonstrates that FDS was either used within the range of its validity as described in NUREG-1824 or that the use of FDS outside the verification and validation range in NUREG-1824 is justified.

Additional Documentation Required Staff needs to review the report of the sensitivity study mentioned on page 11 of the draft response to confirm that it is acceptable for the Radial Distance Relative to Fire Diameter to be outside the validated range. The sensitivity analysis report needs to be posted on the portal.

f. Attachment J of the Transition Report states that the verification and validation of the following applications of Consolidated Model of Fire and Smoke Transport (CFAST) are documented in NUREG-1824.

• HGL Calculations using CFAST (Version 6)

• Temperature Sensitive Equipment Hot Gas Layer Study using CFAST

• Control Room Abandonment Calculation using CFAST Page 5 of 136 to ULNRC-05876 Please provide technical documentation that demonstrates that CFAST was either used within the range of its validity as described in NUREG-1824 or that the use of CFAST outside the verification and validation range in NUREG-1824 is justified.

In addition, please explain why the HGL Calculations using the CFAST calculation described on page J-6 of the Transition Report were not listed as one of the fire models utilized in the application in Section 4.5.1.2.

Additional Clarification/Documentation Required

• MCR Study: The justification for the flame length ratio normalized parameter is adequate, however, there is a confusing comment about the ACRS recently deciding that this parameter should be based not only on the flame height, but the height of the fire above the floor and that this is different from NUREG 1934. This is confusing, since the latest version of NUREG 1934 does include the base height as well as flame height in this normalized parameter. Review this last justification statement and consider whether it should be amended or stricken from the response.

• In addition, in the justification for the equivalence ratio (natural ventilation) normalized parameter being outside of the validation range for heat release rates higher than 312 kW, it is mentioned that a sensitivity study may be warranted, but none is provided in the response. Since the natural ventilation cases with the highest heat release rate bins seem to be the worst-case in terms of the calculated evacuation times, it is requested that this sensitivity study be provided.

Provide the results of the sensitivity study related to equivalence ratio on the portal or, if those results are already posted, indicate where they can be found.

g. During the audit, the NRC staff observed that part of the fire modeling performed in support of the transition to NFPA 805 is described in Engineering Planning & Management, Inc. (EPM)

Report No. R1984-001-002, "Callaway Plant Verification and Validation of Fire Modeling Tools and Approaches." Appendices B, C, and D of this report describe FDS and CFAST fire modeling studies of plume/HGL interaction, temperature sensitive equipment zone of influence (ZOI) and HGL effects. Please provide the basis of assurance that the use of the conclusions from these studies in subsequent fire modeling analysis was within the limits of applicability.

Statement and Justification Needed If the results of the studies described in Appendices B, C and D of the V&V report were used in the analyses within their limits of applicability, a statement should be provided indicating so.

If not, a statement should be provided indicating why not. In addition, the plume-HGL interaction study in Appendix B is based on calculations for a single ambient temperature (70°F), heat release rate (211 kW) and fire size (physical dimensions) and height. Provide justification for drawing generalized conclusions based solely on calculations for these input parameter values.

Page 6 of 136 to ULNRC-05876 Additional Justification Needed Provide a list of areas and scenarios where the results of the analyses described in Appendix B, C and D of document EPM R1984-01-002 were used.

h. Section 4.5.1.2 of the Transition Report lists "Multi-Compartment Analysis Hot Gas Layer Analysis" as one of the fire models utilized in the application. However, there is no verification and validation basis provided for this model in Attachment J. Please explain where this fire model was utilized in the application (if applicable) and provide technical details to demonstrate that the model has been applied within its validated range or to justify the application of the model outside its validated range.

i. During the audit, the NRC staff observed that part of the fire modeling performed in support of transition NFPA 805 is described in EPM Report No. R1984-001-001, "Fire Dynamics Simulator (FDS) Analysis R0." Section C21.3.5 of this report describes how the smoke detector characteristics are prescribed based on Cleary's obscuration correlation.

Please provide the basis for verification and validation of this obscuration correlation. Please provide technical documentation that demonstrates that FDS was either used within the range of its validity as described in NUREG-1824 or that the use of FDS outside the verification and validation range in NUREG-1824 is justified. In addition, please explain why this particular calculation was not listed in Section 4.5.1.2 or Attachment J of the Transition Report.

Additional Justification Needed Reference is made in the draft response to NIST GCR 07-911 for V&V of Clearys smoke detector algorithm, which is now implemented in FDS. The validation in this NIST report is based on three sets of experiments: UL 217 Tests, Room- Corridor-Room Fire Tests and NIST Performance of Home Smoke Alarms Test Validation.

i. UL 217 Tests: Page 24 of the NIST report states that the smoke (soot) yield of polystyrene was estimated as the average of the values in the literature for polystyrene and styrene.

Ignoring the fact that the use of the average in itself is questionable, the SFPE Handbook gives soot yields for polystyrene and styrene as 0.135 and 0.177 g/g, respectively, leading an average of 0.156 g/g. This is approximately three times what was used in the analysis for fire area C-21.

ii. Room-Corridor-Room Fire Test Validation: These experiments consisted of pool fire experiments with a mix of 75% heptane and 25% toluene. The SFPE Handbook lists soot yields for heptane and toluene (listed directly after benzene, but the draft response indicated it was not available) as 0.037 and 0.178 g/g, respectively. The validated soot yield used in these experiments was approximately 0.072 g/g, which is a little bit higher than the value used in the C-21 analysis (0.05 g/g).

Page 7 of 136 to ULNRC-05876 iii. NIST Performance of Home Smoke Alarms Test Validation: As mentioned in the draft response, there is no validated range in this study that can be compared to the analysis in C-21.

In addition, the draft response has some discussion about the literature values of soot yield for polystyrene mattresses and gives the range of 0.056-0.227 g/g. However, the low and high limits of this range are for polyethylene and polyurethane, not polystyrene. The SFPE Handbook gives a range of soot yield for polystyrene foam of 0.18 - 0.21 g/g, which is much higher than was used in the analysis for fire area C-21. Provide additional justification for the validity of Clearys algorithm implemented in FDS for a smoke yield of 0.05 g/g. In addition, explain how the use of a soot yield of 0.05 g/g is justified.

Provide the fuel composition that was used in the FDS analyses of areas C-21 and C-22 and explain how this composition was determined.

j. During the audit, the NRC staff observed that the software package PyroSim (Version 2010.1.0928) was used to build the FDS input files. Please provide technical documentation that demonstrates that PyroSim is verified to build the input file correctly.

Additional Documentation Required.

The draft response states that there is a discussion in the FDS report (R1984-001-001) about the verification of PyroSim. However, PyroSim is not mentioned in the report that is on the portal. Provide the revised FDS report on the portal that discusses PyroSim or explain this discrepancy.

Response to Fire Modeling RAI-01 a) Response provided by ULNRC-05851 dated April 17, 2012.

b) In most cases, the subject correlations have been applied within the validated range reported in NUREG-1824, Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications, Final Report, April 2007. In cases where the models have been applied outside the validated range reported in NUREG-1824, these have been justified as acceptable, either by qualitative analysis, or by quantitative sensitivity analysis. Technical details demonstrating the models are within range, as well as any justification of models outside the range, have been documented in Report No. R1984-001-002, "Verification and Validation of Fire Modeling Tools and Approaches for Use in NFPA 805 and Fire PRA Applications,".

c) Refer to the individual bulleted response items below:

• Heat Detection Actuation Correlation The NUREG-1805, Fire Dynamics Tools: Quantitative Fire Hazard Analysis Methods for the US Nuclear Regulatory Commission Fire Protection Inspection Program, December 2004, heat detection actuation correlation was not considered for fire modeling at Page 8 of 136 to ULNRC-05876 Callaway Plant. There are limited quantities of heat detectors installed at Callaway Plant.

Only those heat detectors near the Turbine Generator were credited in the detailed fire modeling. These heat detectors were credited as part of the pre-action suppression system, which is required for the use of the conditional probability of a catastrophic Turbine Generator scenario found in the guidance of NUREG/CR-6850, Fire PRA Methodology for Nuclear Power Facilities, Final Report, September 2005, Appendix O. In following this guidance, suppression and detection timing are not required and, therefore, the heat detection actuation correlation was not considered. This correlation has been deleted from Section 4.5.1.2 and Attachment J of the Transition Report was revised as shown in Attachment 1 and Attachment J to this enclosure.

• Sprinkler Activation Correlation The Sprinkler Activation Correlation uses the Alpert ceiling jet correlation in addition to a correlation that accounts for the time required to heat the thermal link of the sprinkler. The Alpert ceiling jet correlation is validated in NUREG-1824, Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications, Final Report, April 2007. In most cases, the sprinkler activation correlation has been applied within the validated range reported in NUREG-1824. In cases where the correlation has been applied outside the validated range reported in NUREG-1824, it has been justified as acceptable, either by qualitative analysis, or by quantitative sensitivity analysis. Technical details demonstrating the correlation is used within the validated range, as well as any justification of models outside the range, have been documented in Report No. R1984-001-002, "Verification and Validation of Fire Modeling Tools and Approaches for Use in NFPA 805 and Fire PRA Applications,".

• Corner and Wall Heat Release Rate The Corner and Wall Heat Release Rate correlation is applied within the validated range reported in the studies of Zukoski1, Sargent2, Cetegen3 and Williamson4 or has been justified as acceptable by qualitative analysis or quantitative sensitivity analysis. Technical details have been documented in Report No. R1984-001-002, "Verification and Validation of Fire Modeling Tools and Approaches for Use in NFPA 805 and Fire PRA Applications,".

1 Zukoski, E.E., Properties of Fire Plumes, Combustion Fundamentals of Fire, Cox, G., Ed., Academic Press, London, 1995.

2 Sargent, W.S., Natural Convection Flows and Associated Heat Transfer Processes in Room Fires, Ph.D.

thesis, California Institute of Technology, Pasadena, CA, 1983.

3 Cetegen, B.M., Entrainment and Flame Geometry of Fire Plumes, Ph.D. thesis, California Institute of Technology, Pasadena, CA, 1982.

4 Williamson, R.B. Revenaugh, A. and Mowrer, F.W., Ignition Sources in Room Fire Tests and Some Implications for Flame Spread Evaluation, International Association of Fire Safety Science, Proceedings of the Third International Symposium, New York, pp. 657-666, 1991.

Page 9 of 136 to ULNRC-05876

• Correlation for Heat Release Rates of Cables (Method of Lee)

The Correlation for Heat Release Rates of Cables (Method of Lee) is applied to configurations similar to those reported in NBISR 85-3195, Heat Release Rate Characteristics of Some Combustible Fuel Sources in Nuclear Power Plants, July 1985 or has been justified as acceptable by qualitative analysis. Technical details have been documented in Report No. R1984-001-002, "Verification and Validation of Fire Modeling Tools and Approaches for Use in NFPA 805 and Fire PRA Applications,".

• Correlation for Flame Spread over Horizontal Cable Trays (FLASH-CAT)

The Correlation for Flame Spread over Horizontal Cable Trays (FLASH-CAT) model is applied to configurations similar to those reported in NUREG/CR-7010, Cable Heat Release, Ignition, and Spread in Tray Installations During Fire (CHRISTIFIRE), Draft Report for Comment, September 2010 or has been justified as acceptable by qualitative analysis. Technical details have been documented in Report No. R1984-001-002, "

Verification and Validation of Fire Modeling Tools and Approaches for Use in NFPA 805 and Fire PRA Applications,".

d) The Heskestad plume radius correlation used for fire modeling at Callaway Plant is the same that is used in EPRI fire model FIVE-Rev1, and that is documented in the SFPE Handbook of Fire Protection Engineering, 4th edition, Section 2, Chapter 1. This correlation was not specifically verified and validated (V&Vd) in NUREG-1824, however, Page 2-7 of the 4th edition of the SFPE Handbook of Fire Protection Engineering states that the value calculated by this correlation is the point where the temperature has declined to half of the centerline plume temperature. The Heskestad centerline plume correlation was V&Vd in NUREG-1824.

The plume radius correlation was used in Engineering, Planning and Management Inc. (EPM)

Fire Modeling Database (FMDB) to approximate when to apply the vertical fire plume zone of influence (ZOI), versus the horizontal heat flux based ZOI. The plume radius was not used as the sole basis for any target failures, nor was it used to estimate target temperature. In other words, targets located within the plume radius were considered to be exposed to the centerline temperatures of the plume, while targets located beyond the plume radius were considered to be exposed to the heat flux as determined by the point source model.

Based on how the plume radius was applied, and since the plume radius correlation is a derivative of the Heskestad centerline plume temperature correlation, which was V&Vd by NUREG-1824, the plume radius correlation is subject to the same validated ranges which are described in Callaway Fire Modeling RAI 01-b.

e) In most cases, the FDS analyses have been utilized within the validated range reported in NUREG-1824, Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications, Final Report, April 2007. In cases where the models have been applied outside the validated range reported in NUREG-1824, these have been justified as acceptable, either by qualitative analysis, or by quantitative sensitivity analysis. Technical details demonstrating Page 10 of 136 to ULNRC-05876 the models are within range, as well as any justification of models outside the range, have been documented in R1984-001-002, "Verification and Validation of Fire Modeling Tools and Approaches for Use in NFPA 805 and Fire PRA Applications," Revision 1.

f) HGL Calculations using CFAST (Version 6)

Hot gas layer calculations at Callaway Plant were not performed using CFAST. Hot gas layer calculations were performed using the NUREG-1805, Fire Dynamics Tools: Quantitative Fire Hazard Analysis Methods for the US Nuclear Regulatory Commission Fire Protection Inspection Program, December 2004 and the National Institute of Standards and Technology (NIST) Fire Dynamics Simulator (FDS).

The HGL Calculations using CFAST model has been deleted from Attachment J of the Transition Report as shown in Attachment J to this enclosure.

Control Room Abandonment Calculation using CFAST and Temperature Sensitive Equipment Hot Gas Layer Study using CFAST In most cases, the subject models have been applied within the validated range reported in NUREG-1824, Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications, Final Report, April 2007. In cases where the models have been applied outside the validated range reported in NUREG-1824, these have been justified as acceptable, either by qualitative analysis, or by quantitative sensitivity analysis.

Technical details demonstrating the Temperature Sensitive Equipment Hot Gas Layer Study is within the validated range, as well as any justification of models outside the range, have been documented in Report No. R1984-001-002, " Verification and Validation of Fire Modeling Tools and Approaches for Use in NFPA 805 and Fire PRA Applications,".

Technical details demonstrating the Control Room abandonment models are within range, as well as any justification of models outside the range, will be documented in the next update to Callaway Plant calculation 17671-010b, Callaway NFPA 805 Fire PRA - Main Control Room Fire Analysis per Implementation Item 12-805-002 as shown in the updated Attachment S to this enclosure.

g) The conclusions from the studies in Appendices B, C, and D of R1984-001-002, Callaway Plant Verification and Validation of Fire Modeling Tools and Approaches were used within the limits of applicability established within the study. A detailed discussion of each appendix is provided below:

The conclusions from the plume/HGL interaction study in Appendix B of R1984-001-002 were applied in the fire modeling analysis by correlating each modeled fire compartment to a generic test category analyzed in Appendix B. The correlation was made between fire compartment parameters (i.e., compartment volume and ceiling height), with consideration of fire scenario characteristics (i.e., heat release rate, fire size and elevation). Fire compartments Page 11 of 136 to ULNRC-05876 with parameters within the limits of a generic test category were judged to perform similarly with respect to potential plume and hot gas layer interaction. Section 7.7.1, Basis for Selection of Fire Modeling Tools, within each fire compartment-specific detailed fire modeling report, discusses the details of how the results of the plume/HGL interaction study were applied.

The objective of the Appendix B study was to establish a baseline for whether limitations of using the Heskestad centerline plume temperature correlation calculations, with respect to HGL interaction, need to be accounted for. As such, certain parameters were held constant and this is justified as follows:

• Ambient Temperature. The study aims to find a deviation between the estimated plume temperature with and without a hot gas layer. This deviation is strictly a delta temperature.

Since the delta temperature is the item of interest, the ambient temperature selected is not relevant to the analysis.

• Heat Release Rate. The heat release rate selected represents the majority of fixed ignition sources at Callaway Plant. For those fire compartments where a larger initiating fire is possible or where secondary combustibles may be impacted, the fire scenarios were evaluated, more specifically, to ensure that the target damage set selected bounds any plume/HGL effects.

• Fire Size and Elevation. The fire elevation selected represents a typical electrical cabinet, and bounds the majority of fixed ignition sources at Callaway Plant. For those ignition sources with a fire size (physical dimensions) or elevation exceeding that which was analyzed in Appendix B, the fire scenarios were evaluated, more specifically, to ensure that the target damage set selected bounds any plume/HGL effects.

Page 12 of 136

Enclosure 1 to ULNRC-05876 Examples of the application of Appendix B are as follows:

Zone Zone Appendix Treatment of Fire Sources in the Model Fire Fire Height Volume B Area Zone (ft) (ft3) Category Fixed Ignition Source Fires Transient Fires All targets to the ceiling are considered Category damaged. Therefore the plume-HGL A-8 1316 13 2,080 There are no fixed ignition sources in this fire zone III interaction is bounded by the target damage set.

Category Fire scenarios damage all targets to the ceiling; therefore the plume-HGL interaction is bounded by the TB-1 3705 12 4,800 III target damage set.

For all ignition sources that are able to damage targets (i.e., All targets to the ceiling are considered not well sealed and robustly secured), target damage is damaged. Therefore the plume-HGL A-22 1512 14 41,580 Category II taken to the ceiling. Therefore, the plume-HGL interaction interaction is bounded by the target damage is bounded by the target set. set.

This fire zone is open to Fire Compartment TB-1. Due to the large volume in this fire area, no hot gas TB-1 3619 14 14,336 Category II layer will form.

Fires in this room are bounded by (1) the large open The transient fire size postulated in this geometry that limits plume/HGL interaction, (2) targets room is a 69kW fire. Transient scenarios damaged to the ceiling, (3) a HRR less than 55kW and/or are not large enough to impact secondary (4) the scenario represents suppression activation prior to combustibles. Due to the large size of the C-9 3301 15.2 47,630 Category I 20 minutes credited to prevent damage to fully enclosed room and the small initiating fire, there will cable trays only. be no hot gas layer and no plume/HGL interaction caused by the transient scenarios.

This large fire zone is also open to Fire Compartment TB-1. Due to the large volume in this fire area, no TB-1 4351 17.7 173,184 Category I hot gas layer will form.

The conclusions from the temperature sensitive equipment zone of influence (ZOI) study in Appendix C of R1984-001-002 were applied to the radiant heat ZOI of temperature sensitive components housed in vented metal cabinets. Field conditions were verified to ensure the temperature sensitive equipment met the assumptions and limits of the study (i.e., similar cabinet dimensions and construction, minimum venting requirements at the top and bottom of the cabinet). Section 8.3, Impacts on Sensitive Equipment, within each fire compartment-specific detailed fire modeling report, discusses the details of how the results of the temperature sensitive equipment zone of influence study were applied.

Examples of the application of Appendix C are as follows:

Transient Scenarios D-1.TS-06, D-1.TS-07 and D-1.TS-08 were postulated to surround panels KJ121, NE107 and NG03D, respectively. The panels were confirmed to be vented via field walkdowns. The transient foot print was developed by adding 3-ft along the length and width of the panel and subtracting the percentage of the area occupied by the equipment. This transient footprints are illustrated in Attachment 3 to Calculation KC-75 Detailed Fire Modeling Report for Fire Compartment D-1, Revision 0. The calculation of the transient footprint area is shown in Attachment 7 to KC-75.

The conclusions from the temperature sensitive equipment hot gas layer study in Appendix D of R1984-001-002 were applied in the fire modeling analysis by correlating each modeled fire compartment to a generic test category analyzed in Appendix D. The correlation was made between Page 13 of 136 to ULNRC-05876 fire compartment parameters (i.e., compartment volume and ceiling height), with consideration of fire scenario characteristics (i.e., heat release rate and fire growth profile). Fire compartments with parameters within the limits of a generic test category were judged to perform similarly with respect to gas layer formation. Section 8.3, Impacts on Sensitive Equipment, within each fire compartment-specific detailed fire modeling report, discusses the details of how the results of the temperature sensitive equipment hot gas layer study were applied.

Examples of the application of Appendix D are as follows:

Zone Zone Appendix Fire Fire Height Area D Area Zone (ft) (ft2) Category Application of Appendix D Based on the configuration of the room, there is a possibility for the postulated fires to C-16 3409 15 354 IV generate a hot gas layer capable of damaging all equipment and cable targets in the room. The analysis therefore assumes whole room damage for all fire scenarios in this fire zone.

This fire zone contains no Fire PRA equipment targets and is open to Fire Area A-1.

A-1 1102 24 1000 III Therefore, there is no impact of hot gas layer on temperature sensitive equipment for the fire scenarios in this fire zone.

Based on the results of the study, a hot gas layer is not expected to descend to the height of the plant equipment. Only the tops of the equipment in the area are likely to be exposed and any C-9 3301 15 3139 II equipment likely to contain sensitive components (e.g., switchgear) is well vented. Therefore, there is no impact of hot gas layer on temperature sensitive equipment for the fire scenarios in this fire zone.

Although this fire zone contains multiple electrical panels that may contain temperature TB-1 4401 31 47,524 I sensitive equipment, the large ceiling height and volume of the area prevents the gas layer height from descending to the equipment heights. Therefore, there is no impact of hot gas layer on temperature sensitive equipment for the fire scenarios in this fire zone.

h) Response provided by ULNRC-05851 dated April 17, 2012.

i) The ability of National Institute of Standards and Technology (NIST) Fire Dynamics Simulator (FDS) to determine smoke detector response using the Cleary correlation was not verified and validated in NUREG-1824. However, the ability of FDS to predict detector activation has been verified and validated by Combustion Science and Engineering (CSE), Inc.

and NIST. The analysis of Fire Compartment C-21 (also applicable to Fire Compartment C-

22) has been evaluated against these studies and the model was determined to be appropriate for use in the analysis. Technical details demonstrating the applicability of the validation studies have been documented in Report No. R1984-001-002, " Verification and Validation of Fire Modeling Tools and Approaches for Use in NFPA 805 and Fire PRA Applications".

The use of Fire Dynamics Simulator to predict detector activation has been added to Section 4.5.1.2 and Attachment J to the Transition Report as shown in Attachment 1 and Attachment J to this enclosure. Discussion of the use of Fire Dynamics Simulator within the validated range of NUREG-1824 is contained in the response to FM RAI 01-e.

j) The developers of PyroSim (Thunderhead Engineering) confirmed that PyroSim is verified to build the input file correctly. A multi-level process is used to do this, including testing during Page 14 of 136 to ULNRC-05876 development and running example problems through the software to ensure the correct input data is written and results obtained. Selected examples from NUREG-1824, Verification &

Validation of Selected Fire Models for Nuclear Power Plant Applications, Volume 7: Fire Dynamics Simulator, are used for some of these example problems to ensure the input is written correctly. In addition, PyroSim has been in use by hundreds of users since 2006 and any discrepancies identified by these users are addressed in subsequent releases of the software. Details documenting how PyroSim is verified to build the input file correctly have been added to Report No. R1984-001-002, "Verification and Validation of Fire Modeling Tools and Approaches for Use in NFPA 805 and Fire PRA Applications," Revision 1.

Information on the use of PyroSim to generate the input files has been added to Report No.

R1984-001-001, Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling, Revision 0.

Page 15 of 136 to ULNRC-05876 Fire Modeling RAI 02 NFPA 805, Section 2.7.3.5, "Uncertainty Analysis," states: "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 uncertainty analyses were performed as required by Section 2.7.3.5 of NFPA 805 and the results were considered in the context of the application.

a. Please explain in detail the uncertainty analyses for fire modeling that was performed. Please describe how the uncertainties of the input parameters (geometry, Heat Release Rate (HRR),

Response Time Index (RTI), etc.) were determined and accounted for and substantiate the statement in Appendix J of the Transition Report that states, "... the predictions are deemed to be within the bounds of experimental uncertainty ..."

Additional Justification Needed.

The intent of this RAI was for the licensee to discuss all of the relevant model output parameters described in Table 3-1 of NUREG 1824 (Table 4-1 of NUREG 1934), not just these three. Expand the response to include and justify (as applicable) all the relevant model output parameters described in Table 3-1 of NUREG 1824.

b. During the audit, the NRC staff reviewed EPM Report No. R1984-001-001, "Fire Dynamics Simulator Analysis R0." The staff noted that cable tray obstructions were omitted in the FDS fire modeling analysis for Fire Areas C-21 and C-22.

In a typical fire risk assessment, there are completeness uncertainties in the risk contribution due to scenarios not explicitly modeled (e.g., smoke damage), model uncertainties in the assessment of those scenarios that are explicitly modeled (e.g., uncertainties in the effect of obstructions in a plume), and parameter uncertainties regarding the true values of the model parameters (e.g., the mass burning rate of the source fuel). Please justify why cable tray obstructions could be omitted in the FDS fire modeling analysis for Fire Areas C-21 and C-22.

Additional Justification Needed Provide additional explanation of why hot gas and smoke movement is not affected. During the audit walkdown, NRC staff observed numerous cable trays and other obstructions directly above and adjacent to the ignition source but still within the beam pocket. The calculations assume three cable trays stacked vertically above the fire source location. Provide a detailed explanation of why the additional obstructions are not expected to break up and delay the development of the plume/HGL necessary to activate a detector and/or sprinkler in the beam pocket. In addition, provide a description of the decision process and criteria used for omitting specific obstructions from the FDS analyses.

The results appear to be counter-intuitive and it may be necessary to determine up to what extent the revised geometry in the FDS input file is representative of that in the plant. Provide the FDS input files and pictures that show the obstructions in areas C-21 and C-22.

Page 16 of 136 to ULNRC-05876 Response to Fire Modeling RAI-02 a) The following supersedes the response to this RAI provided by ULNRC-05851 dated April 17, 2012.

Fire modeling in support of the transition has been performed within the Fire PRA, utilizing codes and standards developed by industry and NRC staff which have been verified and validated in authoritative publications, such as NUREG 1824, Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications, Final Report, dated April 2007.

In general, the fire modeling in support of the Fire Risk Evaluations has been performed using conservative methods and input parameters that are based upon NUREG/CR- 6850, Fire PRA Methodology for Nuclear Power Facilities, Final Report, September 2005. This pragmatic approach is used given the current state of knowledge regarding the uncertainties related to the application of the fire modeling tools and associated input parameters for specific plant configurations. A characterization of uncertainties associated with detailed fire modeling has been documented in Section 9 of each fire zone-specific Detailed Fire Modeling Report and is summarized below:

The detailed fire modeling task develops a probabilistic output in the form of target failure probabilities and are subject to both aleatory and epistemic uncertainty.

Appendix V of NUREG/CR-6850 suggests that to the extent possible, modeling parameters should be expressed as probability distributions and propagated through the analysis to arrive at target failure probability distributions. These distributions should be based on the variation of experimental results as well as the analysts judgment. In addition, to the extent possible more than one fire model can be applied and probabilities assigned to the outcome which describe the degree of belief that each model is the correct one. The propagation of fire for each non-screened fire source has been described by a fire model (represented by a fire growth event tree) which addresses the specific characteristics of the source and the configuration of secondary combustibles.

Aleatory uncertainties identified within the fire modeling parameters include:

• Detector response reliability and availability

• Automatic suppression system reliability and availability

• Manual suppression reliability with respect to time available Epistemic uncertainties which impact the zone of influence and time to damage range include:

• Heat release rates (peak HRR, time to reach peak, steady burning time, decay time)

• Number of cabinet cable bundles

• Ignition source fire diameter

• Room ventilation conditions

• Sprinkler Response Time Index (RTI), C factor, and activation temperature

• Detector activation temperature, geometry and obscuration activation Page 17 of 136 to ULNRC-05876

• Soot yield

• Fire growth assumptions (cable tray empirical rule set, barrier delay)

• Cable fire spread characteristics for horizontal and vertical trays

• Transient fires (peak HRR, time to reach peak, location factor, detection time)

• Oil fires (spill assumptions)

• Assumed target location

• Target damage threshold criteria

• Manual detection time

• Mean prompt suppression rate

• Manual suppression rate

• Welding and cutting target damage set

• Transient target impacts With respect to the PRA, a quantitative characterization has not been developed as the quantitative results are conservatively biased for key contributors. Rather than developing a quantitative characterization, an alternate estimate of the mean value for CDF and LERF can be estimated to be a factor of 5 to 10 lower than calculated with a 90 percentile range of a factor of 10 on the lower end and 5 on the higher end. Due to the uncertainty with each of these parameters, the fire modeling task has selected conservative values for each.

Fire models should be created with a substantial safety margin. Per NEI 04-02, there is no clear definition of an adequate safety margin. However, it should be sufficiently large so as to bound the uncertainty within a particular calculation or application. The detailed fire modeling calculations provide a list of items that are modeled conservatively and that provide safety margin. Some examples include the following items:

• Fire scenarios involving electrical cabinets (including the electrical split fraction of pump fires) utilize the 98th percentile HRR for the severity factor calculated out to the nearest FPRA target. This is considered conservative.

• The fire elevation in most cases is at the top of the cabinet or pump body. This is considered conservative, since the combustion process will occur where the fuel mixes with oxygen, which is not always at the top of the ignition source.

• The radiant fraction utilized is 0.4. This represents a 33% increase over the normally recommended value of 0.3.

• The convective heat release rate fraction utilized is 0.7. The normally recommended value is between 0.6 and 0.65, and thus the use of 0.7 is conservative.

• For transient fire impacts, a large bounding transient zone assumes all targets within its ZOI are affected by a fire. Time to damage is calculated based on the most severe (closest) target. This is considered conservative, since a transient fire would actually have a much Page 18 of 136 to ULNRC-05876 smaller zone of influence and varying damage times. This approach is implemented to minimize the multitude of transient scenarios to be analyzed.

• For hot gas layer calculations, no equipment or structural steel is credited as a heat sink, since the closed-form correlations used do not account for heat loss to these items.

• Not all cable trays are filled to capacity. By assuming full, this provides conservative estimates of the contribution of cable insulation to the fire and the corresponding time to damage.

• As the fire propagates to secondary combustibles, the fire is conservatively modeled as one single fire using the fire modeling closed-form correlations. The resulting plume temperature estimates used in this analysis are therefore also conservative, since in actuality, the fire would be distributed over a large surface area, and would be less severe at the target location.

• Target damage is assumed to occur when the exposure environment meets or exceeds the damage threshold. No additional time delay due to thermal response is given.

• The fire elevation for transient fires is 2-feet. This is considered conservative since most transient fires are expected to be below this height or even at floor level.

• Oil fires are analyzed as both unconfined and confined spills with 20-minute durations.

Unconfined spills result in large heat release rates, but usually burn for seconds. The oil fires have been conservatively analyzed for 20-minutes to account for the uncertainty in the oil spill size.

• High energy arcing fault scenarios are conservatively assumed to be at peak fire intensity for 20-minutes from time zero, even though the initial arcing fault is expected to consume the contents of the cabinet and burn for only a few minutes.

• Fire brigade intervention is not credited prior to 85-minutes. Fire Brigade drills indicated that typical manual suppression times can be expected to be much less (i.e., 20 minutes).

All of the fire models used at Callaway Plant and listed in Attachment J of the Callaway Plant NFPA 805 Transition Report were evaluated for experimental uncertainty. The degree to which each model falls within or outside of experimental uncertainty is given in NUREG-1824, Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications, Final Report, dated April 2007. Each model is discussed as follows:

Hot Gas Layer Temperature using FDTs The predictive capability of these parameters using FDTs is characterized as YELLOW+

according to Table 3-1 of NUREG-1824.

Page 19 of 136 to ULNRC-05876 A YELLOW+/- characterization is given: "If the first criterion is satisfied and the calculated relative differences are outside the experimental uncertainty but indicate a consistent pattern of model over-prediction or under-prediction, then the model predictive capability is characterized as YELLOW+ for over-prediction, and YELLOW- for under-prediction. The model prediction for the specific attribute may be useful within the ranges of experiments in this study, and as described in Tables 2-4 and 2-5, but the users should use caution when interpreting the results of the model. A complete understanding of model assumptions and scenario applicability to these V&V results is necessary. The model may be used if the grade is YELLOW+ when the user ensures that model over-prediction reflects conservatism. The user must exercise caution when using models with capabilities described as YELLOW+/-.

NUREG 1824, Volume 3, Section 6.1 states that: The FDTs models for HGL temperature capture the appropriate physics and are based on appropriate empirical data. FDTs generally over-predicts HGL temperature, outside of uncertainty. The over-prediction is expected to lead to conservative results and increased safety margin.

Hot Gas Layer Height and Temperature using FDS The predictive capability of these parameters in FDS is characterized as GREEN according to Table 3-1 of NUREG-1824.

A GREEN characterization is given:" If both criteria are satisfied (i.e., the model physics are appropriate for the calculation being made and the calculated relative differences are within or very near experimental uncertainty), then the V&V team concluded that the fire model prediction is accurate for the ranges of experiments in this study, and as described in Tables 2-4 and 2-5. A grade of GREEN indicates the model can be used with confidence to calculate the specific attribute. The user should recognize, however, that the accuracy of the model prediction is still somewhat uncertain and for some attributes, such as smoke concentration and room pressure, these uncertainties may be rather large. It is important to note that a grade of GREEN indicates validation only in the parameter space defined by the test series used in this study; that is, when the model is used within the ranges of the parameters defined by the experiments, it is validated.

The NUREG-1824, Volume 7, Section 6.1 summary states: FDS is suitable for predicting HGL temperature and height, with no specific caveats, in both the room of origin and adjacent rooms. In terms of the ranking system adopted in this report, FDS merits a Green for this category, based onThe FDS predictions of the HGL temperature and height are, with a few exceptions, within experimental uncertainty.

Hot Gas Layer Temperature and Height using CFAST The predictive capability of these parameters in CFAST is characterized as GREEN according to Table 3-1 of NUREG-1824. The GREEN designation is discussed above under the Hot Gas Layer Height and Temperature using FDS heading. Specifically, the GREEN designation was assigned to the CFAST HGL temperature parameter calculated in the fire compartment of origin. Compartments remote from the fire were assigned a yellow designation. Callaway Plant Page 20 of 136 to ULNRC-05876 only used CFAST to determine the HGL temperature in the fire compartment of origin, and therefore Callaway Plant applications of CFAST fall within the GREEN designation.

The NUREG-1824, Volume 5, Section 6.1 summary states: The CFAST predictions of the HGL temperature and height are, with a few exceptions, within or close to experimental uncertainty. The CFAST predictions are typical of those found in other studies where the HGL temperature is typically somewhat over-predicted and HGL height somewhat lower than experimental measurements. These differences are likely attributable to simplifications in the model dealing with mixing between the layers, entrainment in the fire plume, and flow through vents. Still, predictions are mostly within 10% to 20% of experimental measurements.

Ceiling Jet Temperature using Alpert Correlation The predictive capability of this parameter using the Alpert correlation in the fire model FIVE is characterized as YELLOW+ according to Table 3-1 of NUREG-1824. The YELLOW+

designation is discussed above under the Hot Gas Layer Temperature using FDTs heading.

Specifically NUREG-1824, Volume 5, Section 6.2 summary states: The Alpert correlation under-predicts ceiling jet temperatures in compartment fires with an established hot gas layer.

This result is expected because the correlation was developed without considering HGL effects.

The original version of FIVE accounted for HGL effects by adding the ceiling jet and HGL temperature. This practice results in consistent over-predictions of the ceiling jet temperature.

The approach of adding ceiling jet temperatures to the calculated hot gas layer continues to be the recommended method for FIVE-Rev1 users. Based on the above discussion, a classification of Yellow+ is recommended if HGL effects on the ceiling jet temperature are considered using the approach described in the above bullet. The Alpert correlation by itself is not intended to be used in rooms with an established hot gas layer.

The approach of adding the hot gas layer temperature to the ceiling jet temperature was not used for fire modeling at Callaway Plant. The primary application of the ceiling jet correlation at Callaway Plant was the determination of detection and suppression timing, in which the ceiling jet velocity is a sub-model in the analysis. Including the effects of a hot gas layer would result in shorter detection and suppression times, and therefore the use of the ceiling jet correlation at Callaway Plant is considered conservative. The use of the ceiling jet correlation for target damage is almost always bounded by the use of the point source radiation model and is discussed in detail in the response to RAI 01-b.

Plume Temperature using FDTs The predictive capability of this parameter using FDTs is characterized as YELLOW-according to Table 3-1 of NUREG-1824. The YELLOW- designation is discussed above under the Hot Gas Layer Temperature using FDTs heading.

The NUREG-1824, Volume 3, Section 6.2 summary states: The FDTs model for plume temperature is based on appropriate empirical data and is physically appropriate. FDTs generally under-predicts plume temperature, outside of uncertainty, because of the effects of the hot gas layer on test measurements of plume temperature. The FDTs model is not appropriate for predicting the plume temperatures at elevations within a hot gas layer.

Page 21 of 136 to ULNRC-05876 The use of the FDTs plume correlation for fire modeling applications at Callaway Plant was used within the limitations given in NUREG-1824. The effects of a the plume and hot gas layer interaction were analyzed and documented in detail in Appendix B of R1984-001-002, Verification and Validation of Fire Modeling Tools and Approaches for Use in NFPA 805 and Fire PRA Applications. The use of the FDTs plume correlation was used in accordance with the results of this analysis.

Plume Temperature using FDS The predictive capability of this parameter using FDS is characterized as YELLOW according to Table 3-1 of NUREG-1824.

A YELLOW characterization is given:" If the first criterion is satisfied and the calculated relative differences are outside experimental uncertainty with no consistent pattern of over- or under-prediction, then the model predictive capability is characterized as YELLOW. A YELLOW classification is also used despite a consistent pattern of under- or over-prediction if the experimental data set is limited. Caution should be exercised when using a fire model for predicting these attributes. In this case, the user is referred to the details related to the experimental conditions and validation results documented in Volumes 2 through 6. The user is advised to review and understand the model assumptions and inputs, as well as the conditions and results to determine and justify the appropriateness of the model prediction to the fire scenario for which it is being used.

The NUREG-1824, Volume 7, Section 6.3 summary states: The FDS hydrodynamic solver is well-suited for this application. FDS over-predicts the lower plume temperature in BE #2 because it over-predicts the flame height. FDS predicts the FM/SNL plume temperature to within experimental uncertainty. The simulations of BE #2 and the FM/SNL series are the most time-consuming of all six test series, mainly because of the need for a fairly fine numerical grid near the plume. It is important that a user understand that considerable computation time may be necessary to well-resolve temperatures within the fire plume. Even with a relatively fine grid, it is still challenging to accurately predict plume temperatures, especially in the fire itself or just above the flame tip. There are only nine plume temperature measurements in the data set. A more definitive conclusion about the accuracy of FDS in predicting plume temperature would require more experimental data.

Per the guidance given in NUREG-1934, a D*/x ratio of 5 to 10 produces favorable FDS results at moderate computational cost. This guidance was used for the two Callaway Plant FDS studies that analyzed plume temperatures. The first is the plume and hot gas layer interaction study found in Appendix B of R1984-001-002, Verification and Validation of Fire Modeling Tools and Approaches for Use in NFPA 805 and Fire PRA Applications and the second is an analysis of suppression timing in Compartment C-31 found in Appendix C31 of R1984-001-001, Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling, Revision 0. The D*/x ratio for the critical mesh used in each study is 5.13 and 6.04, respectively, ensuring that the mesh is fine enough to analyze plume temperatures in each case. In addition, the plume temperatures within the flaming region are not the focal point of either study.

Page 22 of 136 to ULNRC-05876 Flame Height using FDTs The predictive capability of this parameter using FDTs is characterized as GREEN according to Table 3-1 of NUREG-1824. The GREEN designation is discussed above under the Hot Gas Layer Height and Temperature using FDS heading.

The NUREG-1824, Volume 5, Section 6.3 summary states: The FDTs model predicted flame heights consistent with visual test observations.

Smoke Concentration using CFAST The predictive capability of this parameter in CFAST is characterized as YELLOW according to Table 3-1 of NUREG-1824. The YELLOW designation is discussed above under the Plume Temperature using FDS heading.

The NUREG-1824, Volume 5, Section 6.6 summary states: CFAST is capable of transporting smoke throughout a compartment, assuming that the production rate is known and its transport properties are comparable to gaseous exhaust products. CFAST typically over-predicts the smoke concentration in all of the BE #3 tests, with the exception of Test 17. Predicted concentrations for open-door tests are within experimental uncertainties, but those for closed-door tests are far higher. No firm conclusions can be drawn from this single data set. The measurements in the closed-door experiments are inconsistent with basic conservation of mass arguments, or there is a fundamental change in the combustion process as the fire becomes oxygen-starved.

Smoke concentration was analyzed in Calculation 17671-010b, Callaway NFPA 805 Fire PRA - Main Control Room Fire Analysis, Revision 1, which was used to determine the probability of Main Control Room evacuation at Callaway Plant following a fire scenario. The Main Control Room is fully enclosed and separated from adjacent areas by vertical control boards and wallboard partitions and was therefore modeled with closed doors. The over-prediction of smoke concentration for closed-door tests as indicated in NUREG-1824 is expected to result in conservative results for this analysis. The smoke production rates used in the model are known and were derived from Table 3-4.16 of the SFPE Handbook of Fire Protection Engineering, 4th Edition. Transport properties of the smoke are expected to be comparable to gaseous exhaust products.

Oxygen Concentration using CFAST The predictive capability of this parameter in CFAST is characterized as GREEN according to Table 3-1 of NUREG-1824. The GREEN designation is discussed above under the Hot Gas Layer Height and Temperature using FDS heading.

Page 23 of 136 to ULNRC-05876 The NUREG-1824, Volume 5, section 6.5 summary states: CFAST uses a simple user-specified combustion chemistry scheme based on a prescribed pyrolysis rate and species yields that is appropriate for the applications studied. CFAST predicts the major gas species close to experimental uncertainty.

Radiant Heat using FDTs The predictive capability of this parameter in FDTs is characterized as YELLOW according to Table 3-1 of NUREG-1824. The YELLOW designation is discussed above under the Plume Temperature using FDS heading.

The NUREG-1824, Volume 3, Section 6.4 summary states: The FDTs point source radiation and solid flame radiation model in general are based on appropriate empirical data and is physically appropriate with consideration of the simplifying assumptions. The FDTs point source radiation and solid flame radiation model are not valid for elevations within a hot gas layer. FDTs predictions had no clear trend. The model under- and over-predicted, outside uncertainty. The point source radiation model is intended for predicting radiation from flames in an unobstructed and smoke-clear path between flames and targets.

Only the FDTs point source radiation model was used for fire modeling at Callaway Plant.

NUREG- 1824 indicates that there is no clear trend in under or over-prediction for the point source model. The model over-predicted heat flux for locations immersed in a hot gas layer, which is likely due to smoke and the HGL preventing radiation from reaching the gauges. This over-prediction is expected to lead to conservative results and increased safety margin. In a smaller number of cases, the model under-predicted heat flux due to contribution of radiation from the HGL. In order to account for this potential under-prediction, conservatism has been built into the use of the radiation model at Callaway Plant, including the use of a radiant heat release rate fraction of 0.4.

In addition, NUREG-1824 indicates that the point source model is not intended to be used for locations relatively close to the fire. For fire modeling at Callaway Plant, targets located close to the fire have conservatively been failed within the early stages of fire growth.

Radiant Heat using FDS The predictive capability of this parameter in FDS is characterized as YELLOW according to Table 3-1 of NUREG-1824. The YELLOW designation is discussed above under the Plume Temperature using FDS heading.

Even though the FDS Radiant Heat Model was given a Yellow designation, NUREG 1824, Volume 7, Section 6.8 states that: FDS has the appropriate radiation and solid phase models for predicting the radiative and convective heat flux to targets, assuming the targets are relatively simple in shape. FDS is capable of predicting the surface temperature of a target, assuming that its shape is relatively simple and its composition fairly uniform. FDS predictions of heat flux and surface temperature are generally within experimental uncertainty, but there are numerous exceptions attributable to a variety of reasons. The accuracy of the predictions generally decreases as the targets move closer to, or go inside of, the fire. There is not enough near-field data to challenge the model in this regard.

Page 24 of 136 to ULNRC-05876 FDS was used to calculate radiant heat exposure at Callaway Plant for two applications. The first application was to determine the radiant heat exposure to an electrical cabinet from a transient fire. The second application was to determine the heat flux levels at potential targets from a transient fire. For both applications, the limitations outlined in NUREG 1824 are not of concern because:

1) Heat flux is not being calculated for any targets inside of the fire. For both FDS analyses performed, all potential radiant heat targets are located a minimum of 3 feet horizontally away from the fire.

2) All targets are simple in shape and not complex in nature. The targets analyzed in the two FDS models are a flat sheet metal panel and heat flux monitoring devices located independently from obstructions. In both instances, the targets are of simple geometry and uniform composition.

Since the model was not used outside of the limitations identified, it is concluded that the FDS predictions of heat flux is within experimental uncertainty.

b) Although the FDS model did not exactly replicate the field conditions in terms of cable tray obstructions, a sensitivity study determined that omitting these does not significantly affect the output parameters being evaluated in the FDS model (i.e., automatic detection and suppression system activation). The detection and suppression timing determined in the sensitivity study does not change the target damage set determined for the scenarios. The analysis is documented in Sections C21.7.2 and C22.7.2 of the FDS Report R1984-001-001, Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling. The FDS computer input file is provided in Enclosure 2.

Page 25 of 136 to ULNRC-05876 Fire Modeling RAI 03 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. The NRC staff noted that no specific discussion was found in the Transition Report, with respect to how the input for the algebraic models were established for fires that involved multiple combustibles. Please explain how the input for the algebraic models was established for fires that involved multiple combustibles and justify the approach that was used.

b. The NRC staff noted an apparent lack of specific discussion in the Transition Report regarding how the input for the CFAST models was established for the Main Control Room (MCR) evacuation study. Please describe the specific CFAST input parameters and provide the CFAST input files for the MCR evacuation study.

c. During the audit, the NRC staff noted that fire modeling report R1984-001-001 "Fire Dynamic Simulator Analysis to Support Detailed Fire Modeling," Rev. O. states in several places (all Appendices) that, "the mesh size reflects the finest mesh feasibly allowable with the given computer resources." Please explain why the mesh size used is within the validated range and confirm whether a grid sensitivity study was performed or justify why such a study was not performed.

Additional Documentation Required:

The draft response states that a sensitivity study was conducted for Mesh 1 in Fire Area C-1 to show that results of a 0.05-m mesh would yield the same conclusions as the 0.1-m mesh used in the original analysis. Provide the grid sensitivity study on the portal.

d. During the audit, the NRC staff noted that Section A11.1 of fire modeling report, R1984-001-001 discussed how the analysis performed for Fire Area C-31 was applied to Fire Area A-11 since the room dimensions for both spaces are comparable. However, this discussion does not describe how the ignition source location and the radial distance between the fire source and the sprinkler was selected.

Please explain how the assumption to use the FDS analysis for Fire Area C-31 to apply to Fire Areas A-11 and C-30 is adequate. In addition, please explain how the ignition source location and secondary combustibles in Fire Areas A-11 and C-30 are considered by the analysis of Fire Area C-31.

e. During the audit, the NRC staff noted fire modeling report R1984-001-001 states "It should be noted that NUREG 1824 did not provide verification and validation for estimating sprinkler activation times. However, the major inputs used in the determination of suppression Page 26 of 136 to ULNRC-05876 (determination of gas temperatures) have been validated." Based on this statement, it was not clear to the staff how the sprinkler activation time was determined. Please explain how the sprinkler activation time was calculated in the FDS analysis.

Additional Documentation Required The Heskestad/Bill equation and part of the text appear to be missing from the non-docketed section. Provide the missing information in the response.

f. During the audit, the NRC staff noted that different material properties were used in the FMDB analysis as in the FDS analysis for the same fire areas (A-11, C-21, etc.). For example. in Calculation No. KC-49, the material properties used in the FDS analysis for concrete is different from that used in the FMDB and transient datasheet analysis. The thermal conductivity and density in the FMDB are 1.6 Watts per meter Kelvin (W/m-K) and 2400 kilograms per cubic meter (kg/m3) as opposed to 1.0 W/m-K and 2100 kg/m3 used in FDS. The specific heat of concrete in FDS calculations is 0.88 kilojoules per kilogram Kelvin (kJ/kg-K) and in FMDB calculations are 0.75 kJ/kg-K.

Please explain the reason for the difference in material properties used in FMDB and FDS analyses. In addition. please explain what effect the difference in material properties used in the analyses has on the conclusions.

Additional Clarification Required The licensee stated that the difference in material properties does not significantly affect the thermal inertia and therefore this difference will not affect the results of the analysis. Provide justification for the statement the difference in the thermal inertia values is not significant.

g. During the audit, the NRC staff noted that fire modeling report R1984-001-001 discussed how the water discharge spray is input into FDS for each sprinkler head and there are figures in each Appendix that show water spray from an activated sprinkler. Based on this discussion, it was not clear to the staff how the sprinkler water spray characteristics were used in the FDS analysis. Please explain how the sprinkler water spray characteristics were used in the FDS analysis.

h. During the audit, the NRC staff noted that Section A11.3.5.1 of fire modeling report R1984-001-001 discussed why the heat release rate profile was chosen instead of:

1. A smaller initial fire size which, along with ignition of secondary combustibles might result in quicker sprinkler activation, or

2. A larger initial ignition source which would not activate sprinklers prior to ignition of secondary combustibles.

Page 27 of 136 to ULNRC-05876 Based on this discussion in the report, it was not clear to the NRC staff how these assumptions were verified. Please explain how the heat release rate profiles chosen were conservative for the purposes of damage assessment and sprinkler activation. In addition, please apply this response to the analysis conducted for the other two cable chase fire areas (C-30 and C-31) analyzed with FDS.

Additional Justification Needed The licensees approach recognizes the fact that there is a trade-off between choosing a low vs.

a high heat release rate. The former delays detector and sprinkler activation but also results in less damage. To be conservative a relatively high peak heat release rate (317 kW) and a relatively slow growth rate (8 min to peak heat release rate) were used. Explain why only these two values were used and why a different set of equally plausible values would not result in greater risk.

i. During the audit, the NRC staff noted that fire modeling report R1984-001-001 stated that a slice temperature file was created at ceiling level to analyze the sprinkler activation times.

Based on this statement, it is not clear to the NRC staff how the sprinkler activation time was determined (slice file output or FDS sprinkler activation algorithm). In this same section of each FDS analysis, there is a discussion about the slice file output showing that the fire ignition location does not affect the results in terms of sprinkler activation. Please explain how the sprinkler activation time is determined in the FDS analysis and provide technical justification for the conclusion that the slice file output shows that fire location does not affect the sprinkler activation times.

Additional Documentation Required Part of the draft response between pages 1 and 2 appears to be missing. Provide the missing information in the response.

j. During the audit, the NRC staff noted that Section C21.2 of fire modeling report R 1984-001-001 states, in part, that, "the purpose of the FDS simulation was to determine the time at which the ceiling-mounted quick-response sprinklers in this fire compartment would activate as a result of a transient fire." However, in the paragraph that follows, it is stated that the sprinklers were given an RTI of 130 milliseconds0.5 (m-s0.5), which is a value more typical of a standard response sprinkler. Please state what type of sprinklers are in the lower Cable Spreading Room (CSR) and also provide a justification for the RTI used in the analysis.

Additional Justification Needed See action for RAI 3k response below.

k. During the audit, the NRC staff noted that Section C21.3.5 of fire modeling report R1984-001-001 states that standard response sprinklers are used in the CSR and therefore an RTI of 130 (m-s)0.5 was used for the analysis. The licensee justified this value for the RTI by way of Page 28 of 136 to ULNRC-05876 reference to NUREG-1805, which provides a generic RTI value of 130 (m-s)0.5 for standard response heads with a fusible link. However, in Chapter 10 of NUREG-1805, there is a note about selecting the RTI of a sprinkler element which states, " the actual RTI should be used when the value is available." Please provide justification for the RTI value chosen for this analysis and describe how that value compares with the RTI of the actual sprinklers in the CSRs. In addition, please apply the response to the upper CSR (Fire Area C-22).

Additional Justification Needed In the FDS analysis of fire area C-21, a sprinkler head RTI of 130 (ms)0.5 was used based on NUREG-1805. This value was justified because standard response sprinklers have an RTI of 80(m-s)1/2 or higher and the use of 130(m-s)1/2 is therefore conservative. The draft response refers to a NIST study, which is cited as the basis of the default values in NUREG 1805.

However, the objective of that study was to compare four sprinkler activation models. Another study by the same author and published in Operation of Fire Protection Systems (special addition to the NFPA Handbook), shows that the typical range for standard response sprinkler RTIs is approximately 100-350(m-s)1/2. Provide justification for the use of an RTI of 130(m-s)1/2 in lieu of the more conservative values reported in the literature.

A typical range for standard response sprinkler response time index (RTIs) is 100-350 (m-s)1/2.

Perform a sensitivity analysis to substantiate the use of an RTI of 130 (m-s) 1/2.

I. During the audit, the NRC staff noted that Section C21.3.5.1 of fire modeling report R 1984-001-001 discusses why a 45 kilowatt (kW) initiating fire was considered more conservative than a 69 kW initiating fire, in terms of sprinkler activation and ignition of secondary combustibles. It was not clear to the staff how this conservatism was verified. Please explain how heat release rate profiles chosen were conservative for the purposes of damage assessment and sprinkler activation. In addition, please apply this response to the analysis conducted for the other upper CSR (C-22) analyzed with FDS.

m. During the audit, the NRC staff noted that Section C21.4 of R1984-001-001 of fire modeling report states, in part, that" ... is expected to result in suppression activation within 13.5 minutes.

This timing directly corresponds to ignition of the third cable tray in a stack." In Section C21.3.5.1 of the report, it was stated that, "The third cable tray ignites at 12 minutes." This language suggests that the third cable tray ignites at the same time as sprinkler activation.

Please clarify what is meant by this statement and how the ignition of the third cable tray affects the sprinkler activation time.

n. During the audit, the NRC staff noted Section C21.5 of fire modeling report R1984-001-001, states that "The modeled configuration of a transient fire in C-21 does not result in the formation of a hot gas layer before automatic suppression is actuated." Please provide technical justification for this statement. In addition, please apply this response to the analysis conducted for the other upper CSR (C-22) analyzed with FDS.

Page 29 of 136 to ULNRC-05876

o. During the audit, the NRC staff noted that Section C21.5 of fire modeling report R 1984-001-001 states that "The FDS analysis results for Fire Compartment C-22 are based on the analysis performed for Fire Area C-21, the lower CSR. The C-21 analysis results for suppression activation are considered equivalent to those expected in C-22 due to their similar configurations." However, the ceiling of C-21 is specified as approximately 25 feet and the ceiling of C-22 is specified as approximately 12 feet, respectively. Please explain this difference in ceiling height and why it was not necessary to model C-22 separately.

p. During the audit, the NRC staff reviewed Scientech Calculation 17671-010b, "Callaway NFPA 805 Fire PRA Main Control Room Fire Analysis," and discussed the analysis with the licensee.

During this discussion, NRC was told that it was assumed that a fire originating in the Equipment Cabinet Area (ECA) was assumed to not be able to propagate into the MCR. Please provide a basis for this assumption.

Additional Information Needed The concern also included the possibility of HGL forming above the MCR cabinets, as they did not extend to the ceiling and also any other means of fire spread via the open ceiling area, not just cabinet-to-cabinet directly. Exposed cables, if any, could also be targets along which fire could propagate between the two areas.. The draft response discusses why direct cabinet-to-cabinet fire propagation is precluded, but does not discuss other potential modes of fire spread.

Provide a discussion of other potential modes of fire spread in the response.

q. During the audit, the NRC staff reviewed Scientech Calculation 17671-010b, and discussed the analysis with the licensee's staff. During this discussion, it was stated that it was assumed that there was only qualified cable in the MCR. However, Section 2 of Attachment 1 (Control Room Evacuation Study) of this calculation states that it is assumed the control room contains both qualified and unqualified cabling. Please clarify whether there is unqualified cable in the control room and if so, what is the ratio of unqualified to qualified cable.

r. During the audit, the NRC staff reviewed Scientech Calculation 17671-010b, and discussed the analysis with the licensee. In Table 1 of Attachment 1 (Control Room Evacuation Study) of Scientech Calculation 17671-010b, the modeled fire scenarios are provided. For single cabinet fires, both qualified and unqualified cabling was used in the calculation of evacuation times.

However, for the multicabinet fire scenarios, only qualified cable was considered in the calculation of evacuation times. Please explain why unqualified cable was not considered for multi-cabinet fires.

s. During the audit, the NRC staff reviewed Scientech Calculation 17671-101b, as well as Attachment 1 (Control Room Evacuation Study). The fuel combustion properties for qualified and unqualified cable are provided in this report. The heat of combustion (HOC) for qualified and unqualified cable is given as 28.3 and 20.9 megajoules per kilogram (MJ/kg), respectively.

It is not expected that the HOC for an unqualified cable would be lower than a qualified cable.

Please confirm these material values and also explain how the HOC material property is used in the analysis.

Page 30 of 136 to ULNRC-05876

t. Please provide the FDS input files for the detailed FDS fire modeling conducted as described in EPM document Nos. R1984-001-001 and R1984-001-C1, Detailed Fire Modeling Report -

FDS Analysis of HDPE Pipes (Draft B).

Additional Information Needed The Society of Fire Protection Engineers (SFPE) Handbook lists two sets of values for the type of qualified cable that is present in the MCR (XLPE/XLPE according to the MCR abandonment study report). XLPE/XLPE cable #1 in the SFPE handbook table has a soot yield of 0.12 g/g and a heat of combustion of 28.3 kJ/g. The values for cable #2 are 0.12 g/g and 12.5 kJ/g, respectively. Explain why the soot yield and heat of combustion for cable #1 were used in the analysis.

Response to Fire Modeling RAI-03 a) The approach for fires involving multiple combustibles was to calculate the heat release rate of each individual fire as a function of time, and then use the combined total heat release rate as the input to the algebraic models. Conservative heat release rates were determined from NUREG/CR-6850, and the rules for propagation to cable trays, and fire spread rates all followed the FLASH-CAT model found in NUREG/CR-7010. This approach is considered appropriate for the following reasons:

• The approach is endorsed in Section 3.2.2.2 of NUREG-1934, second draft report for comment, which discusses summing up individual heat release rates for use in algebraic models:

The heat release rate from the cable tray can be added to the heat release rate of the cabinet to determine a combined heat release rate as a function of time. This total rate can then be used in the various models as an approximation of the heat release rate as a function of time.

• Using the sum of all heat release rates is expected to result in conservative estimation of zone of influence as calculated by the algebraic models. In a realistic setting, each individual fire taken separately would create smaller zones of influence than that calculated for one large, combined fire. This is in part due to the expected interference of the base fire on the plume entrainment and flame heights of the secondary combustible fires, resulting in the reduction of the effective mass burning rate of the secondary combustible fires. In addition, the obstructing fires could create an environment where the fire would be oxygen limited.

• The fire diameter used as the input to the algebraic models is equal to the fire diameter of the original source fire and remains unchanged throughout the burning duration of the fire. In reality, a spreading fire will have an increasingly larger fire Page 31 of 136 to ULNRC-05876 diameter. The use of the source diameter is considered more severe for plume and flame height correlations, as the use of a small diameter results in a stronger plume and thus larger vertical zone of influence values.

• Burnout was considered; however, spread along cable trays was modeled until 85 minutes. NUREG/CR-7010, Section 9.2.4 states that the assumption that the fire will spread laterally until all cable is consumed is conservative, as this phenomenon was not observed in many of the multiple tray experiments. Assuming total consumption of all cables leads to conservative heat release rates and zone of influence calculations.

b) The Callaway Plant Main Control Room (MCR) evacuation study used Consolidated Model of Fire Growth and Smoke Transport (CFAST) models as described in Attachment A to calculation 17671-10b. A summary of the plant drawings and plant procedures used as inputs are provided below. Fire Modeling RAI-01-f provides additional information about parameter selection and validation of usage. The CFAST model input files for the Callaway Plant MCR are provided in Enclosure 2.

• Control Building HVAC System design basis description document, ULDBD-GK-001, Rev. 0

• Control Room Inaccessibility procedure, OTO-ZZ-00001, Rev. 32

• Computer Room & Control Room Detailed Floor Plans, EL 2047-6, Drawing A-2337 Rev. 11

• Heating, Ventilation & Air Cond. Control Building EL. 2047-6, Drawing M-2H3611 Rev. 7

• Heating, Ventilation & Air Cond. Control Building Sections & Details, Drawing M-2H3901 Rev. 6

• Hot Lab, Counting Room, Control Room & Computer Room Reflected Ceiling Plans, Drawing A-2332 Rev. 9 c) NUREG-1824, Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications, Final Report, April 2007, defines the validated range as a D*/x value from 4 to 16. Further, NUREG-1934, Nuclear Power Plant Fire Modeling Application Guide, Second Draft for Comment, July 2011, states that D*/x ratios of 5 to 10 usually produce favorable results at a moderate computational cost. Finer meshes would not necessarily provide more accurate results while causing a dramatic increase in computation time.

For each FDS analysis documented in R1984-001-001, Fire Dynamics Simulator (FDS)

Analysis to Support Detailed Fire Modeling, Revision 0, the D* and D*/x values were calculated. The mesh sizes utilized in the models were determined to be either acceptably refined based on industry guidance or a sensitivity study was performed to determine that the results and conclusions were valid. The statement the mesh size reflects the finest mesh size feasibly allowable with the given computer resources has been removed and the results of the D* analysis have been added to Sections A11.4.1, C31.4.1, C30.4.1, C21.4.1 and C22.4.1 of Page 32 of 136 to ULNRC-05876 R1984-001-001 Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling, Revision 0.

d) Response provided by ULNRC-05851 dated April 17, 2012.

e) Sprinkler activation times were predicted by the NIST Fire Dynamic Simulator (FDS) which uses the differential equation of Heskestad and Bill (published in Fire Safety Journal volume 14, 1988) to determine the link temperature. This method calculates sprinkler activation time based on the gas temperature and the sprinkler parameters. The sprinkler activation time is, therefore, reliant on the ability of FDS to predict ceiling jet and gas layer temperatures. NUREG-1824, Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications, Final Report, April 2007, assigned FDS a Green ranking for its ability to predict both ceiling jet and gas layer temperatures.

Based on the validation of the FDS models ability to predict ceiling jet/gas temperature, and that the Heskestad and Bill method is documented in an authoritative publication, the use of FDS to determine sprinkler activation time is considered acceptable. This information has incorporated into Appendix H of Report R1984-001-002, Verification and Validation of Fire Modeling Tools and Approaches for Use in NFPA 805 and Fire PRA Applications, Revision 1.

f) The Fire Dynamics Tools (FDTs) presented in NUREG 1805, Fire Dynamics Tools:

Quantitative Fire Hazard Analysis Methods for the US Nuclear Regulatory Commission Fire Protection Inspection Program, December 2004, suggest thermal properties for common boundary materials. The Fire Dynamics Simulator (FDS) developed by the National Institute of Standards and Technology (NIST), comes equipped with a material properties database that also contains the thermal properties for a range of common materials. The Callaway Plant fire models used the thermal properties as suggested by the specific tool as the values for the installed boundary materials.

The Fire Modeling Database (FMDB) utilizes the equations provided by the NUREG 1805 FDTs and the FMDB results have been verified and validated against those generated by the NUREG 1805 spreadsheets. Therefore, the calculations performed using the FMDB used the suggested material properties provided in NUREG 1805 FDTs 2.1, 2.2 and 2.3.

The material properties used in the NIST FDS models are taken from the material properties database provided by NIST for use in FDS. The material properties in the database have been reviewed against the material properties found in the SFPE Handbook of Fire Protection Engineering, 4th Edition, and match the SFPE suggested values.

The NUREG 1824, Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications, Final Report, April 2007, FDS verification and validation study includes the following discussion on the selection of material properties:

Some of the property data needed by FDS are commonly available in fire protection engineering and materials handbooks. Depending on the application, properties for specific Page 33 of 136 to ULNRC-05876 materials may not be readily available (especially burning behavior at different heat fluxes). A small file distributed with the FDS software contains a database with thermal properties of common materials. This data are given as examples, and users should verify the accuracy and appropriateness of the data.

The inputs used in the FDS models described in R1984-001-001, Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling, Revision 0 employed this methodology by selecting the material properties from the FDS database and validating the parameters against the suggested values in the SFPE Handbook, 4th edition. Therefore, the use of these values is considered appropriate.

g) Response provided by ULNRC-05851 dated April 17, 2012.

h) Sprinkler activation and time to target damage are dependent on the heat release rate (HRR) of the fire and the temperatures generated at the sprinkler/target location. A larger heat release rate will result in quicker, more severe target damage, but will also prompt suppression to activate earlier in the scenario. Consequently, a conservative heat release rate for target damage is a non-conservative heat release rate for sprinkler activation. Therefore, in order to avoid non-conservative assumptions for target damage or suppression activation, a conservative heat release rate was employed for both analyses. A detailed discussion is included in R1984-001-001, Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling, Revision 0.

i) The slice files shown in R1984-001-001, Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling, Revision 0 were included in the report to demonstrate that the gas temperatures at the ceiling are relatively uniform. Although only the slice file at the time of activation is included in the report, ceiling temperatures were observed to be uniform throughout the Smokeview simulation. Based on the even heating of the upper gas layer, the model is not sensitive to minor variations in the sprinkler location with respect to the fire location selected. Since a bounding fire size and location were chosen and the gas layer is expected to be of uniform temperature, all sprinkler locations with respect to the fire location have been bounded by the analysis. The discussion in R1984-001-001 has been clarified to provide additional discussion of the use of the slice file.

j) Contrary to Section C21.2 of fire modeling report R1984-001-001, Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling, Revision 0, the cable spreading rooms are not equipped with quick-response sprinklers. The installed ceiling-mounted sprinklers are standard response type. Fire Modeling Report R1984-001-001 has been revised to correct the error.

The sprinkler head in use in fire areas C-21 and C-22 is a Star Model E, 165 °F, Spray Nozzle that uses a fusible link. Star Sprinkler Co. is no longer in business and the specific Response Time Index (RTI) for the Model E is unknown. Standard response sprinklers have a RTI value of 80 m*s or more as defined by NFPA 13 2002 edition. Because the specific RTI is unknown the NUREG-1805, Fire Dynamics Tools: Quantitative Fire Hazard Analysis Page 34 of 136 to ULNRC-05876 Methods for the US Nuclear Regulatory Commission Fire Protection Inspection Program, December 2004 FDT Section 10.3.3 generic RTI of 130m*s for standard response sprinklers with a thermal link was utilized. The RTI is a laboratory test obtained value that rates the thermal sensitivity of a sprinkler and the RTI is used primarily to categorize a sprinkler head as either Quick Response or Standard Response. In field installed conditions numerous factors affect actual sprinkler activation times including sprinkler temperature ratings, ceiling height, fire distance to the sprinkler, fire heat release rates, and sprinkler head distance from the ceiling. For C-21 and C-22 sprinkler activation is assumed after three trays are damaged or the equivalent of 14 minutes when following the EPRI cable tray fire propagation guidance. This sprinkler activation time is realistic and acceptable for use in the FPRA analysis.

k) Contrary to Section C21.3.5 of fire modeling report R1984-001-001, the cable spreading rooms are not equipped with quick-response sprinklers. The installed ceiling-mounted sprinklers are standard response type. Fire Modeling Report R1984-001-001 has been revised to correct the error.

The sprinkler heads used in fire areas C-21 and C-22 are a Star Model E, 165 degree F, spray nozzle that uses a fusible link. Star Sprinkler Company is no longer in business and the specific Response Time Index (RTI) for the Model E spray nozzle is unknown. Standard response sprinklers have a Response Time Index (RTI) value of 80 (m-s)1/2 or more as defined by NFPA 13, 2002 edition. Because the specific RTI is unknown, the NUREG-1805 FDT Section 10.3.3 generic RTI of 130 (m-s)1/2 for standard response sprinklers with a thermal link was utilized. The RTI value is obtained from laboratory testing which rates the thermal sensitivity of a sprinkler. The RTI is used primarily to categorize a sprinkler head as either Quick Response or Standard Response. In field installed conditions numerous factors affect actual sprinkler activation times including sprinkler temperature ratings, ceiling height, fire distance to the sprinkler, fire heat release rates, and sprinkler head distance from the ceiling.

The FDS analysis for C-21 and C-22 was performed only to determine the time to suppression activation. Fire propagation in the cable trays was determined using the timing of NUREG/CR-6850 Appendix R. Based on this guidance, cable tray ignition occurs as shown below:

Cable Tray Ignition Time (minutes)

Tray #1 5 Tray #2 9 Tray #3 12 Tray #4 14 Tray #5 15 Tray #6 16 The FDS analysis determined that sprinkler activation will occur prior to ignition of cable tray

1. 4, which occurs at 14 minutes.

The detailed fire modeling task develops a probabilistic output in the form of target failure probabilities that are subject to both aleatory and epistemic uncertainty. Appendix V of Page 35 of 136 to ULNRC-05876 NUREG/CR-6850 suggests that to the extent possible modeling parameters should be expressed as probability distributions and propagated through the analysis to arrive at target failure probability distributions. These distributions should be based on the variation of experimental results as well as the analysts judgment. In addition, to the extent possible more than one fire model can be applied and probabilities assigned to the outcome which describe the degree of belief that each model is the correct one.

The propagation of fire is described by a fire model (represented by a fire growth event tree),

which addresses the specific characteristics of the source and the configuration of secondary combustibles.

Aleatory uncertainties identified within the fire modeling parameters include:

• Detector response reliability and availability

• Automatic suppression system reliability and availability

• Manual suppression reliability with respect to time available Epistemic uncertainties which impact the zone of influence and time to damage range include:

• Heat release rates (peak HRR, time to reach peak, steady burning time, decay time)

• Ignition source fire diameter

• Room ventilation conditions

• Fire growth assumptions (cable tray empirical rule set, barrier delay)

• Cable fire spread characteristics for horizontal and vertical trays

• Transient fires (peak HRR, time to reach peak, location factor, detection time)

• Assumed target location

• Target damage threshold criteria

• Manual detection time

• Mean prompt suppression rate

• Manual suppression rate

• Transient target impacts Due to the uncertainty with each of these parameters, the fire modeling has selected conservative values for each as discussed below.

• Transient fires were postulated to occur throughout the compartment even though, based on the configuration, not all areas are accessible or realistic locations for transient ignition sources.

• For transient fire impacts, a large bounding transient zone assumes all targets within its footprint and within the zone of influence are affected by fire. This is considered conservative, since a transient fire would actually have a much smaller zone of influence.

• The HRR of the initial transient fire used was 45 kW verse 69 kW.

Page 36 of 136 to ULNRC-05876

• The transient fire location was assumed to be 2 ft. below the first tray. However, in actual plant configurations most trays are located more than 2 ft. above the assumed fire elevation.

• The radiant fraction is assumed to be 0.4. This represents a 33% safety margin over the normally recommended value of 0.3. In addition, the convective heat release rate fraction utilized is 0.7. The normally recommended value is between 0.6 and 0.65.

• No credit is given in the FDS model for metal cable tray bottoms or covers that would delay damage or fire growth. Some cable trays have tray bottoms and covers installed.

• The HRR used in the FDS model does not include cable tray risers. Where cable trays risers exist that are adjacent to horizontal trays in the damage assessment they are assumed damaged however for the HRR used in FDS they are not included because they significantly increased HRR and reduced sprinkler activation time.

• The transient fire location was chosen based on a worst case (farthest) distance from the sprinklers, in order to ensure the most conservative delayed suppression activation was modeled.

• Target damage is assumed to occur when the exposure environment meets or exceeds the damage threshold.

• When suppression fails, whole room damage is always assumed to occur. This is considered conservative as the damage set and time to damage would be less severe.

• Fire brigade intervention is not credited prior to 85-minutes. Fire Brigade drills indicated that typical manual suppression times can be expected to be much less.

A sensitivity analysis was performed to determine the sprinkler activation times assuming only the RTI value is changed over the range of RTI values allowed for a standard response sprinkler which is 80 to 350 (m-s)1/2. As previously noted, numerous factors affect actual sprinkler activation times including sprinkler temperature ratings, ceiling height, fire distance to the sprinkler, fire heat release rates, and sprinkler head distance from the ceiling.

Page 37 of 136 to ULNRC--05876 ACTIIVATION DEELTA TO A ASSUMED RTI TIME E ACCTIVATION N 80 12 miin 33 sec - 877 sec 130 13 miin 34 sec - 255 sec (-3%)

165 13 miin 52 sec - 077 sec (< -1%

%)

225 14 miin 17 sec + 1 8 sec (2%)

350 15 miin 28 sec + 889 sec Sincce Star Sprinnkler is no longer in busin ness and thee design of thhe Star Moddel E sprinkleer pred dates industry y-wide use of o the RTI teest, a specificc RTI value is not availaable. Howevver, the geneeral range off RTI values for the sprin nkler may bee establishedd based on exxperience wwith simiilar design spprinklers. Th he range of typical t RTI values for a standard ressponse sprinnkler was discussed with w a represeentative from m Factory M Mutual. Factoory Mutual iss responsiblee for the standdard FM 200 00, Approv val Standard for Automat atic Control MMode Sprinkklers for Firee Protection. Thee representattive from Facctory Mutuaal indicated tthat the Star Model E sprrinkler headd design utiliizes a fusiblee plug which h would not be expectedd to fall withiin the lowest 1/2 portiion of the allowable rang ge (80-120) (m-s) . It w was also noteed an RTI vaalue over 225 (m-1/2 s) is i unlikely fo or the sprink kler. Therefo ore, based onn feedback frrom Factoryy Mutual, thee geneeral range off RTI for thee Star Model E sprinkler would reasoonably be exxpected to faall in the 1/2 rang ge of 130-225 (m-s) baased on its deesign. As shhown in the ttable, the sprrinkler activvation timee assumed in n the fire mod deling for firre areas C-21 and C-22 ffall within thhe range of tthe Pagge 38 of 136 to ULNRC-05876 expected RTI values. The epistemic uncertainty due to any one input such as the sprinkler RTI is addressed by the fire modeling selecting conservative values for a number of the inputs to 1/2 account for that uncertainty. The sprinkler RTI of 130 (m-s) is within the expected range of RTI for a standard response sprinkler with the Star Model E design based on discussions with 1/2 an industry expert. The use of a sprinkler RTI value of 130 (m-s) results in a sprinkler activation time that is realistic and acceptable for use in the FPRA analysis. Uncertainty in the sprinkler RTI value is addressed similar to other epistemic uncertainty by use of conservative parameters in other fire model inputs.

l) Sprinkler activation and time to target damage are mainly dependent on the heat release rate (HRR) of the fire and the temperatures generated at the sprinkler/target location. A larger heat release rate will result in quicker, more severe target damage, but will also prompt suppression to activate earlier in the scenario. Consequently, a conservative heat release rate for target damage is a non-conservative heat release rate for sprinkler activation. In order to avoid non-conservative assumptions for target damage or suppression activation, a different heat release rate was employed for both analyses.

The cable spreading rooms contain a large number of horizontal cable tray stacks. The timing of fire propagation between cable tray stacks is prescribed by NUREG/CR-6850, Fire PRA Methodology for Nuclear Power Facilities, Final Report, September 2005 and the analysis follows this timing. Therefore, a fire that is quickly suppressed will be prevented from spreading into multiple trays and reduce the number of targets damaged by the spreading fire and expanding zone of influence. The fire suppression analysis in R1984-001-001, Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling, Revision 0 was constructed to ensure a conservative estimate of suppression activation. This critical fire size was determined to be 45kW, as cable trays are located 2-ft above the floor. A fire generating less than 45kW will impact only floor-based targets and will not cause ignition of the tray stacks, and would, therefore, be non-conservative with respect to damage.

However, selecting a fire larger than 45kW will result in increased gas layer temperatures. As Fire Dynamics Simulator (FDS) uses the temperature of the gas layer to determine the link/bulb temperature of the sprinkler to predict activation, increasing the HRR and gas temperature will cause suppression to activate earlier in the scenario. Therefore, in order to maximize fire growth and spread without promoting suppression, the fire suppression analysis in R1984-001-001 used a 45kW fire instead of a 69kW fire.

The fire suppression analysis in R1984-001-001 is used only to determine the time to suppression activation given the prescribed fire scenario. The fire growth analysis is performed in the detailed fire modeling calculations KC-68, Detailed Fire Modeling Report for Fire Compartment C-21 and KC-69, Detailed Fire Modeling Report for Fire Compartment C-22.

The zone of influence for target damage was determined using a 69kW transient fire. The use of the larger initiating fire results in a conservative target damage set. The fire growth analyses Page 39 of 136 to ULNRC-05876 and target damage sets are documented in Detailed Fire Modeling Calculations KC-68 and KC-69.

m) The statement this timing corresponds to ignition of the third cable tray in a stack is meant to describe that suppression activates after the third cable try ignites, based on the combined heat release rate of the three trays. The FDS analysis contained in R1984-001-001, Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling, Revision 0 was performed to determine the time to suppression activation given the fire growth analysis developed in Calculation KC-68, "Detailed Fire Modeling Report for Fire Compartment C-21". Timing of cable tray ignition was determined using the empirical rule-set prescribed in NUREG/CR-6850, Fire PRA Methodology for Nuclear Power Facilities, Final Report, September 2005 Appendix R.4.2. Since cable tray ignition time is not determined by the FDS analysis, the statement relating cable tray ignition time to suppression timing is not appropriate for inclusion in R1984-001-001 and has been removed.

n) The statement the modeled configuration of a transient fire in C-21 (and applicable to C-22) does not result in the formation of a hot gas layer before automatic suppression is actuated has been removed from sections C21.5 and C22.5 of Report R1984-001-001, Fire Dynamic Simulator (FDS) Analysis to Support Detailed Fire Modeling, Revision 0. The FDS analysis was performed to evaluate suppression activation time only and not to evaluate hot gas layer formation within the compartment. Therefore, this statement is not required in the report.

o) Response provided by ULNRC-05851 dated April 17, 2012.

p) Propagation of fire from the back panels in the electrical cabinet areas (ECA) to the main control board (MCB)was not considered due to the configuration of the panels in the ECA.

Callaway Plant drawing J-24001 shows the configuration of the ECA. The back wall of the main control board (RL028 to RL012) is a solid steel plate. It has no fire protection rating, but it also has no openings. The distance between the RP053 cabinets and the back of the main control board is approximately 5 feet. That is the same distance between the SB32, SB29 cabinets and the main control boards RL018-RL012. The distance from the single cabinet RP068 to the main control board is about 3 feet, but RP068 is a single bundle cabinet.

Appendix S.1 of NUREG/CR-6850 states cabinet to cabinet propagation can be ignored if cabinets are separated by a double wall with an air gap. This is the case for the ECA back panels and the back of the main control board. This supports the engineering judgment to not consider propagation from the back panels in the ECA to the main control boards.

Page 40 of 136 to ULNRC-05876 Other Fire Spread Mechanisms:

Horizontal Propagation via cable trays: The ECA has a ceiling height of 26 ft. Heat and combustion products from panel fires in the ECA will most likely rise straight up to the ceiling.

Some of the panels in the ECA have cable trays exiting the top. All but 3 of the cable trays that exit the panels rise directly up to the upper cable spreading room (which is above the Main Control Room(MCR)). For panels with vertical cable trays exiting the top, fire propagation in the horizontal direction to the main control board is highly unlikely. There are 3 trays that originate from panel RK045E and cross over to the MCA (main control area) over the top of the MCB panels, over the MCR acoustical ceiling. The RK045E panels are more than 20 ft.

from the main control board. Drawings show the bottom of the cable trays from RK045E are 9 ft. over the cabinets they pass over. The cables are IEEE rated so growth is postulated to be very slow. Horizontal propagation via cable trays was dismissed based on engineering judgment given the geometry and material properties.

Hot gas layer: The probability of a hot gas layer from the ECA causing damage to the main control board is bounded by the probability that a hot gas layer causes evacuation of the MCR.

The CFAST calculation provides scenarios for HGL formation in the MCA from fires in the ECA. ECA fires causing damage to MCB and NOT causing evacuation are not seen as a separate scenario.

q) All cable in the MCR cabinets is qualified to IEEE-383 standards. The MCR evacuation study evaluated qualified and unqualified cable fires as part of a project requirement. At the time the MCR evacuation study was performed in 2009, the cable content of the MCR had not been verified. The scope of the evacuation study was set to be applicable to all eventualities. The runs for unqualified cable are not used in the analysis.

r) All cable in the MCR cabinets is qualified to IEEE-383 standards. When the MCR evacuation study was first performed in 2009, the cable content of the MCR had not been verified. The scope of the evacuation study was set to be applicable to all eventualities. The MCR evacuation study evaluated qualified and unqualified cable fires as part of a project requirement. In 2009, only single cabinet scenarios were considered.

The MCR evacuation study was updated in 2011, to account for multi-cabinet fires. By that time, it had been verified that all cable was qualified, so there was no need to perform a multi cabinet run with unqualified cable.

s) The heat of combustion for qualified and unqualified table was derived from table 3-4.16 of the SFPE Handbook of Fire Protection Engineering, 4th edition. They are different from what one would expect intuitively. The Callaway Plant Fire PRA model is based on qualified cable throughout the plant (with the exception of some areas of the Turbine Building) and is documented as such in the MCR calculation (17671-10b). As a sensitivity case, the CFAST model was also quantified with unqualified cable so both HOC values are shown in the documentation.

Page 41 of 136 to ULNRC-05876 t) The FDS input files are provided in Enclosure 2. The FDS input files have also been added as Appendix I to Report No. R1984-001-001, Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling, The SFPE Handbook Table 3-4.16 gives two sets of data for XPLE/XPLE cables. The first of which gives the HoC as 28.3 MJ/kg which is the value used in the analysis. The second set of data gives the HoC as 12.5 MJ/kg. The soot yield is given as 0.12 g/g in both cases. CFAST gives higher values of optical density with higher HoC values, and therefore shorter times to reach abandonment conditions and higher abandonment probabilities. The higher value for HoC is therefore conservative with respect to the time to abandonment based on optical density.

Page 42 of 136 to ULNRC-05876 Section 2: Response to Fire Protection RAIs Fire Protection Engineering RAI 01 In Attachment A of the LAR, Table B-1, on page A-25, the compliance statement for NFPA 805 Section 3.3.7.1 states "complies with clarification." The compliance basis states: "Bulk hydrogen complies with the requirements of NFPA 50A-1973. Exceptions requiring further action are identified below." Another compliance statement "complies with required action" is used. The compliance basis states "see implementation items identified below." There are two implementation items associated with this requirement.

It is unclear what the clarification is and whether or not the required actions are necessary for the entire chapter 3 attribute. Please clarify the use of the two-part compliance statement and what the clarification is intended to be.

Response to Fire Protection Engineering RAI-01 Response provided by ULNRC-05851 dated April 17, 2012.

Fire Protection Engineering RAI 02 In Attachment A of the LAR, Table B-1, on page A-33, the compliance statement for NFPA 805 Section 3.4.1(a)(1) states "complies with clarification." The compliance basis states "the industrial fire brigade complies with NFPA 600-2000 Edition. Exceptions requiring further action are identified below."

It is unclear what the clarification is and whether or not the required actions are necessary for the entire chapter 3 attribute. Please clarify the use of the two-part compliance statement and what the clarification is intended to be.

Response to Fire Protection Engineering RAI-02 Response provided by ULNRC-05851 dated April 17, 2012.

Page 43 of 136 to ULNRC-05876 Fire Protection Engineering RAI 03 In Attachment A of the LAR, Table B-1, on page A-37, the compliance statement for NFPA 805 Section 3.4.2 states "complies," however, implementation items are listed below. Please clarify whether "complies" is the correct compliance statement with the requirements in this section or if the plant complies with required action or both.

Response to Fire Protection Engineering RAI-03 Response provided by ULNRC-05851 dated April 17, 2012.

Fire Protection Engineering RAI 04 In Attachment A of the LAR, Table 8-1, on page A-39, the compliance statements for NFPA 805 Sections 3.4.2.3 and 3.4.2.4 state "complies, with required action," and the compliance basis states "see implementation item identified below." It was noted that there are no implementation items identified below these two sections. Please identify the required actions.

Response to Fire Protection Engineering RAI-04 Response provided by ULNRC-05851 dated April 17, 2012.

Fire Protection Engineering RAI 05 In Attachment A of the LAR, Table B-1, on page A-45, the compliance statement for NFPA 805 Section 3.4.4 states "complies with clarification." The compliance basis states "Equipment is provided for the fire brigade as required. Per visual inspection of equipment, it is in accordance with applicable NFPA codes, as documented in CAR 200902315." However the clarification is not apparent. Please identify the clarification used to support the compliance statement.

Response to Fire Protection Engineering RAI-05 Response provided by ULNRC-05851 dated April 17, 2012.

Page 44 of 136 to ULNRC-05876 Fire Protection Engineering RAI 06 In Attachment A of the LAR, Table B-1, on page A-59, the requirements of NFPA 805 Section 3.5.15 for fire hydrants and hose houses are stated. The LAR states that the exception to this section in NFPA 805 is utilized which provides a mobile means of providing hose and associated equipment in lieu of hose houses. The exception states the mobile equipment shall be equivalent to the equipment supplied by three hose houses. The compliance basis states that equipment on two mobile units is provided, but does not specify the amount of equipment provided. Please clarify the actual equipment equivalency for the mobile units.

Response to Fire Protection Engineering RAI-06 Response provided by ULNRC-05851 dated April 17, 2012.

Fire Protection Engineering RAI 07 In Attachment A of the LAR, Table B-1, on page A-64, the compliance statement for NFPA 805 Section 3.6.2 states "complies with clarification." However the clarification is not apparent. Please identify the clarification used to support the compliance statement.

Response to Fire Protection Engineering RAI-07 Response provided by ULNRC-05851 dated April 17, 2012.

Fire Protection Engineering RAI 08 In Attachment A of the LAR, Table B-1, on page A-66, the compliance statement for NFPA 805 Section 3.6.4 states "compliance by previous NRC approval." The compliance basis for this element does not address the provision of this section to provide manual fire suppression in areas containing systems and components needed to perform nuclear safety functions following a safe shutdown earthquake. Although not addressed in the LAR, 10 CFR 50.48(c)(vi) states NRC requirements for licensees that wish to apply the exception to Section 3.6.4. Please describe how compliance is achieved with the requirement to provide manual fire suppression to protect nuclear safety functions in the event of a safe shutdown earthquake.

Response to Fire Protection Engineering RAI-08 Based on NRC feedback, the compliance statement for this section is acceptable as stated and no further action is required.

Page 45 of 136 to ULNRC-05876 Fire Protection Engineering RAI 09 In Attachment A of the LAR, Table B-1, on page A-83, the compliance statement for NFPA 805 Section 3.9.3 states "complies with clarification." The compliance basis states that water flow alarms annunciate on panels that connect to KC008, which is located in the control room. Similarly, in Attachment A of the LAR, Table B-1, on page A-89, the compliance statement for NFPA 805 Section 3.10.2 also states "complies with clarification." The compliance basis states that all system actuation alarms annunciate on panels that connect to KC008, which is located in the control room. Please provide further discussion on these clarifications, including a description of the alarm process and how the alarming condition is communicated to the operator(s).

Response to Fire Protection Engineering RAI-09 Response provided by ULNRC-05851 dated April 17, 2012.

Fire Protection Engineering RAI 10 On December 21, 2011, there was a fire in the B emergency diesel generator (EDG) jacket water heater where the breaker for the heater did not automatically open and a fire was reported on the paint on the outside of the heater. Subsequently, the jacket water heater was determined to be non-functional and jacket water temperature dropped below the technical specification (TS) required limit and the B EDG was declared inoperable. Please describe the effects this incident, if any, and any subsequent actions taken as a result of this incident, have on the NFPA 805 LAR and the transition process.

Response to Fire Protection Engineering RAI 10 Response provided by ULNRC-05851 dated April 17, 2012.

Fire Protection Engineering RAI 11 Section 4.1.2.3 and Attachment L, Approval Request 1, of the LAR describe the storage and refilling capacity of the fire protection water storage tanks to demonstrate that the requirement for two separate 300,000 gallons supplies is not adversely impacted by using the fire protection water supply for non-fire protection purposes. Please describe the administrative and/or operating procedures used to ensure that the minimum required fire protection water supply remains available.

Response to Fire Protection Engineering RAI-11 Response provided by ULNRC-05851 dated April 17, 2012.

Page 46 of 136 to ULNRC-05876 Fire Protection Engineering RAI 12 Table B-1, Criteria 3.5.1(b), Fire Flow Rate of the LAR indicates that compliance with this item is not applicable. However, in Approval Request 1, compliance to this requirement, namely, the 500 gallons per minute (gpm) hose stream requirement, is the basis for the request. Please reconcile the discrepancy.

Response to Fire Protection Engineering RAI-12 Response provided by ULNRC-05851 dated April 17, 2012.

Fire Protection Engineering RAI 13 In Attachment A of the LAR, Table B-1, on page A-91, the compliance basis for NFPA 805 Section 3.10.9 does not provide adequate detail to conclude that the possibility of secondary thermal shock damage was considered for the design of the gaseous fire suppression systems at Callaway plant.

Please provide additional information to justify the conclusion that Halon 1301 does not present a risk of secondary thermal shock.

Response to Fire Protection Engineering RAI-13 Additional details have been added to the License Amendment Request (LAR), Transition Report, Table B-1, Section 3.10.9 compliance statement which describe that the Halon systems do not present a risk of secondary thermal shock. The revised Transition Report, Table B-1, Section 3.10.9 is provided in Attachment A to this enclosure.

Fire Protection Engineering RAI 14 NFPA 805, Section 3.9.1 requires that water-based fire suppression systems be installed in accordance with the appropriate NFPA standard. During the audit, it was observed that quick response sprinkler heads were installed in multiple cable chases, replacing the original sprinkler nozzles. Due to the piping configuration, the quick response sprinkler heads were installed at an angle relative to the ceiling, as opposed to being parallel to it; the latter of which is typical.

Plant modification item 201002877 to install the quick response sprinklers in cable chases A-11, C-30, and C-31 has been completed. Please provide the basis and justification for compliance to the appropriate NFPA standard.

Response to Fire Protection Engineering RAI-14 Response provided by ULNRC-05851 dated April 17, 2012.

Page 47 of 136 to ULNRC-05876 Fire Protection Engineering RAI 15 NFPA 805, Section 3.9.1(1) requires that the standpipe systems comply with the NFPA 14, "Standard for the Installation of Standpipe, Private Hydrant, and Hose Systems" code of record (i.e., 1976).

During the audit, the licensee indicated normal working pressures range from 150-160 pounds per square inch (psi). In accordance with NFPA 14, Section 4-4.2, the pressures should not exceed 65 psi for Class I connections (1.5-inch) and 100 psi for Class II connections (2.5-inch). Please provide a description of the system pressures at the hose connections and whether or not these pressures exceed the required values. If pressures exceed these values, please provide the justification and basis for having the higher pressure(s). Include any prior approvals and any justification for meeting any other NFPA 14 requirements, as necessary. Please update the code conformance review calculation document as necessary.

Response to Fire Protection Engineering RAI-15 Response provided by ULNRC-05851 dated April 17, 2012.

Page 48 of 136 to ULNRC-05876 Section 3: Response to Monitoring Program RAIs Monitoring Program RAI 01 NFPA 805, Section 2.6, "Monitoring," states that "a monitoring program shall be established to ensure that the availability and reliability of the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria" and that "Monitoring shall ensure that the assumptions in the engineering analysis remain valid."

Specifically, NFPA 805, Section 2.6 states that 2.6.1 Acceptable levels of availability, reliability, and performance shall be established.

2.6.2 Methods to monitor availability, reliability, and performance shall be established. The methods shall consider the plant operating experience and industry operating experience.

2.6.3 If the established levels of availability, reliability, or performance are not met, appropriate corrective actions to return to the established levels shall be implemented. Monitoring shall be continued to ensure that the corrective actions are effective.

Section 4.6, "Monitoring Program," of the Transition Report states that the NFPA 805 monitoring program will be implemented "after the safety evaluation issuance as part of the fire protection program transition to NFPA 805" (Table S-3, Implementation Items, item 11-805-089 of the Transition Report).

Furthermore, the licensee has committed to comply with Frequently Asked Question (FAQ) 10-0059.

The NRC staff noted that the information provided in Section 4.6, "Monitoring Program," of the Transition Report is insufficient for the staff to complete its review of the monitoring program, and, as such, is requesting that the following additional information be provided.

a. A description of the process by which systems, structures, and components (SSCs) will be identified for inclusion in the NFPA 805 monitoring program, including the approach to be applied to any fire protection SSCs that are already included within the scope of the Maintenance Rule program.

b. A description of the process that will be used to assign availability, reliability, and performance goals to SSCs within the scope of the monitoring program including the approach to be applied to any SSCs for which availability, reliability, and performance goals are not readily quantified.

Page 49 of 136 to ULNRC-05876

c. A demonstration of how the monitoring program will address response to programmatic or training elements that fail to meet performance goals (examples include fire brigade response or performance standards and discrepancies in programmatic areas such as combustible programs).

d. A description of how the monitoring program will address fundamental fire protection program elements.

e. A description of how the guidance in EPRI Technical Report 1006756, "Fire Protection Equipment Surveillance Optimization and Maintenance Guide" will be integrated into the monitoring program.

f. A description of how periodic assessments of the monitoring program will be performed taking into account, where practical, industry wide operating experience including whether this process will include both internal and external assessments and the frequency at which these assessments will be performed.

Response to Monitoring Program RAI-01 The LAR Transition Report Section 4.6.2 has been revised to align with the approved FAQ 10-0059 and its related closure memo. The revised Section 4.6.2 describes the process by which systems, structures, and components (SSCs) will be identified for inclusion in the NFPA 805 monitoring program, including the approach to be applied to any fire protection SSCs that are already included within the scope of the Maintenance Rule program. The revised LAR Transition Report Section 4.6.2 is provided in attachment 1 to this enclosure. Additionally, LAR Transition Report Attachment H and Attachment S have been updated to reflect that FAQ 10-0059 has now been approved.

a. The LAR Transition Report Section 4.6.2 has been revised to align with the approved FAQ 10-0059 and its related closure memo. The revised Section 4.6.2 provides a description of the process that will be used to assign availability, reliability, and performance goals to High Safety Significant (HSS) SSCs within the scope of the monitoring program. Low Safety Significant (LSS) SSCs do not specifically require assignment of availability, reliability, and performance goals. Programmatic elements such as fire brigade performance, fire watches, combustible controls, etc., will be evaluated using the existing program health process. It is not practical to assign target values of reliability and availability to these attributes so their effectiveness is based on objective and anecdotal evidence evaluated by plant personnel in charge of the fire protection programs as is currently practiced. The revised Section 4.6.2 is provided in Attachment 1 to this enclosure.

b. The LAR Transition Report Section 4.6.2 has been revised to align with the approved FAQ 10-0059 and its related closure memo. The revised Section 4.6.2 provides a description of how the monitoring program will address response to programmatic elements that fail to meet performance goals. The revised Section 4.6.2 is provided in Attachment 1 of this enclosure.

c. The LAR Transition Report Section 4.6.2 has been revised to align with the approved FAQ 10-0059 and its related closure memo. The revised Section 4.6.2 provides a description of how Page 50 of 136 to ULNRC-05876 the monitoring program addresses fire protection systems and features and programmatic elements. The revised Section 4.6.2 is provided in Attachment 1 of this enclosure.

d. As identified in the LAR Transition Report, Attachment A, Table B-1, Section 3.2.3.1, the frequency at which inspections, testing and maintenance of the fire protection systems and features are performed will be evaluated using EPRI Technical Report 1006756, "Fire Protection Equipment Surveillance Optimization and Maintenance Guide". EPRI Technical Report 1006756 Section 11 contains the following guidance which ensures that reliability levels established are consistent with the FPRA and Maintenance Rule Program:

In establishing reliability goals, each plant should determine if other programs, evaluations, or analyses have credited specific reliability values. For example, if the Fire PRA credits a specific level of reliability for a certain suppression system, the target reliability for surveillance optimization should not be below the credited value.

e. The LAR Transition Report Section 4.6.2 has been revised to align with the approved FAQ 10-0059 and its related closure memo. The revised Section 4.6.2 provides a description of how periodic assessments of the monitoring program will be performed including consideration of internal and external operating experience. The revised Section 4.6.2 is provided in Attachment 1 of this enclosure.

f. The LAR Transition Report Section 4.6.2 has been revised to align with the approved FAQ 10-0059 and its related closure memo. The revised Section 4.6.2 provides a description of how periodic assessments of the monitoring program will be performed including consideration of internal and external operating experience. The revised Section 4.6.2 is provided in Attachment 1 of this enclosure.

Page 51 of 136 to ULNRC-05876 Section 4: Response to Safe Shutdown RAIs Safe Shutdown Analysis RAI 01 NEI 00-01, "Guidance for Post-Fire Safe Shutdown Circuit Analysis," Revision 1, Alignment -

provide a gap analysis on the differences between the alignments using NEI 00-01, Revision 1, as the basis for transitioning the NFPA Standard 805 nuclear safety capability as indicated in NEI 04-02, "Guidance for Implementing a Risk-informed, Performance Based Fire Protection Program Under 10 CFR 50.48( c)," versus using NEI 00-01, Revision 2, which is the current version cited in Regulatory Guide 1.205, "Risk Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1.

Response to Safe Shutdown Analysis RAI-1 Response provided by ULNRC-05851 dated April 17, 2012.

Safe Shutdown Analysis RAI 02 The nuclear safety capability assessment (NSCA) assumed the loss of instrument air. Please explain how this was incorporated into the initial position of components for circuit analysis. Also, please explain how instrument air failure was considered in the non-power operations (NPO) analysis.

Response to Safe Shutdown Analysis RAI-2 Response provided by ULNRC-05851 dated April 17, 2012.

Safe Shutdown Analysis RAI 03 Section 4.1.2.2 and Attachment T, Clarification Request 1 of the LAR -NUREG-0830, "Safety Evaluation Report Related to the Operation of Callaway Plant, Unit No.1," Supplement 3, states that "Some operations require cutting a control power cable at the equipment to ensure that a fault in the control room does not prevent certain equipment operation." Please explain if these operations are retained in the transition to NFPA 805. If so, please explain how these were considered as variations from the deterministic requirements in the NFPA 805 analysis.

Response to Safe Shutdown Analysis RAI-3 Response provided by ULNRC-05851 dated April 17, 2012.

Page 52 of 136 to ULNRC-05876 Safe Shutdown Analysis RAI 04 Section 4.1.2.2 and Attachment T, Clarification Request 2 of the LAR -Please explain if there are any significant ignition sources or combustible loading in the vicinity of the subject emergency or equipment hatch that can challenge the non-rated penetrations. Please explain if there has been any significant change to the room configuration since previous approval.

Response to Safe Shutdown Analysis RAI-4 Response provided by ULNRC-05851 dated April 17, 2012.

Safe Shutdown Analysis RAI 05 Section 4.1.2.2 and Attachment T, Clarification Request 4 of the LAR -The LAR states that "The original NRC approval was granted based on the overall design of the fire protection features in the rooms and did not specifically rely on the dike capacity." This conflicts with other information provided in the LAR. Please specify the capacity of the diesel fuel oil day tank dike system and justify if the system remains adequate with the reduced capacity of less than 100 percent.

Response to Safe Shutdown Analysis RAI-5 Response provided by ULNRC-05851 dated April 17, 2012.

Page 53 of 136 to ULNRC-05876 Safe Shutdown Analysis RAI 06 Section 4.2.1.2 and Table B-2 of the LAR -To extend the minimum 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> coping time, operators must take action to recharge the nitrogen accumulators to support emergency operation of the atmospheric steam dump (ASD) valves and the turbine drive auxiliary feedwater pump (TDAFW) to steam generator (SG) flow control valves. Please explain if the components and/or cables associated with this action are included in the NSCA safe shutdown (SSD) equipment list. Please explain if the steps for recharging the nitrogen accumulators detailed in plant procedures are demonstrated to be feasible.

Since the actions to recharge the nitrogen accumulators are not considered recovery actions, please provide a qualitative risk analysis that demonstrates that the risk of failing to perform the actions within the required time frame is low.

Should the accumulators not be recharged, please explain if the TDAFW flow control valves can be locally throttled. If so, please explain how these steps are proceduralized and demonstrated to be feasible.

Response to Safe Shutdown Analysis RAI-6 Response provided by ULNRC-05851 dated April 17, 2012.

Page 54 of 136 to ULNRC-05876 Section 5: Response to Probabilistic Risk Assessment RAIs Probabilistic Risk Assessment RAI 01 The disposition of several Facts and Observations (F&Os) for the internal events PRA model identifies that the item is resolved and thereby included in the current internal events model but not incorporated into the FPRA model. During the audit, the licensee identified that the internal events PRA model has been revised since the development of the FPRA and has undergone a focused scope peer review after the fire peer review was completed. Please provide the following:

a. A description of any changes made to the internal events PRA model which are not part of the FPRA and disposition any potential impact on the FPRA results.

b. A description of the focused scope peer review and disposition any F&Os resulting from this review for their applicability to the current FPRA model.

c. A discussion of the overall impact of the changes to the internal events PRA model in terms of how the internal events risk profile has changed, that the changes would not impact the FPRA results, and that the internal events PRA model used in the FPRA development can be considered to represent the as-built and operated plant even though additional changes have subsequently been made to the internal events model.

Additional Justification Needed Even if unaffected by fire, new HFEs/HEPs in the internal events PRA model can still affect the fire PRA results because there may be scenarios initiated by fire where non-fire-affected HFEs/HEPs are part of the mitigation. Therefore, item (iii) is applicable to the fire PRA, but it is possible that all such HFEs/HEPs are re-evaluated in light of potential fire effects such that it was determined that there were no changes. If the latter is correct, revise the statement.

The following are questions on the new F&Os in Table 1 (note that this includes Suggestions, to cover the possibility that a Suggestion relative to the internal events PRA could have a greater impact on the fire PRA):

(1) F&O 1-7. Can rupture of the RHR or SI system be induced by a fire, including a conditional rupture resulting from a fire-induced initiator? If so, how? It appears the F&O cites an underestimate of the rupture probability and, if this is somehow incorporated into the fire PRA, what is the effect on CDF, delta-CDF, LERF and delta-LERF, at least based on a bounding estimate, to ensure the transition conclusions are not affected?

(2) F&O 1-13. If the updated CCF probabilities indicate increases, and these CCFs are part of the fire PRA, what is the effect on CDF, delta-CDF, LERF and delta-LERF, at least based on a bounding estimate, to ensure the transition conclusions are not affected?

Page 55 of 136 to ULNRC-05876 (3) F&O 4-5. Is FW availability after core damage credited in the LERF model for the fire PRA? If so, would not this then be applicable?

The responses need to consider not just the FPRA results for the proposed changes which are part of the LAR, but also consider the requested self-approval after implementation of the NFPA 805 license amendment. If appropriate, in order to justify the existing model, please provide sensitivity studies using the updated internal events conditional core damage probabilities.

Response to PRA RAI-01

a. Response provided by ULNRC-05851 dated April 17, 2012.

b. Response provided by ULNRC-05851 dated April 17, 2012.

c. The internal events risk profile changed with the recent PRA update relative to PRA model revision 4, primarily in that CDF has decreased overall. The three principle reasons are as follows:

i. Addition of a non-safety related motor driven Auxiliary Feedwater Pump ii. The installation of additional offsite power capability from an electrical cooperative substation and the addition of 4 diesel generators from an offsite location iii. Changes to the HRA Items (i) and (ii) were incorporated into the fire PRA (FPRA) prior to submitting the NFPA-805 LAR application. Item (iii) has also been accounted for in the FPRA, because FPRA is required to develop fire specific HEPs, even for the internal events PRA Human Failure Events. As such, the HEPs used in the FPRA are specifically re-evaluated in light of potential fire effects and no updates to the FPRA HEPs were deemed necessary in light of the recent internal events PRA HRA updates.

As noted, the significant changes to the internal events PRA that have caused a decrease in internal events plant risk have already been reflected in the FPRA. As such, no significant changes in the FPRA risk results or insights (i.e. that would affect the NFPA-805 LAR application) would be expected if the full extent of the internal events PRA changes were incorporated into the FPRA at this time.

Table 1 that follows is a summary of the PWROG focused-scope internal events peer review F&Os and their disposition. All F&Os have been addressed, except where noted.

Page 56 of 136 to ULNRC-05876 PRA RAI 1 Table 1 - PWROG Focused-Scope Internal Events Peer Review F&Os and Their Disposition F&O Associated (F)inding or No. SR(s) (S)uggestion Brief Description Disposition for IE PRA Disposition for FPRA 1-1 IC-C5 F Convert SSIE and A plant availability factor of 0.9, FPRA calculated its own IEFs. [n/a]

ISLOCA IE frequencies to used for the other internal events Rx-yr basis. initiators, was applied to the SSIE and ISLOCA frequency quantification.

1-2 IE-C8 F Assess alternate Calculations EA-05 Revision 1 FPRA models multiple alignments for alignments for loss of SW and EG-19 Revision 1 were ESW and CCW and uses split fractions and CCW. modified to provide more in- for percent of time spent in each one.

depth discussion and [n/a]

justification for the use of a single alignment.

1-4 IE-C10 F More than one initiator The ISLOCA and Loss of All Does not apply. FPRA has its own BE exists in SSIE and Service Water models were initiators and calculates its own ISLOCA cutsets. revised such that cutsets now ISLOCA frequency. [n/a]

contain only one initiator/frequency BE.

1-71 IE-C14 F Rupture probability of The RHR and SI system rupture FPRA uses overpressure probabilities RHR and SI systems probabilities used in ZZ-138, from ZZ-138. The updated probabilities needs to be based on Rev. 0, Add. 1 were revised in in ZZ-138, Rev 0, Addenda 1 have not failure probability of all response to this F&O. For both been incorporated into the FPRA as of piping/components (not systems, a summation of the March 2012. These will be updated on the weakest location). piping/component rupture with next FPRA update.

probabilities is now used. Fire-induced rupture [due directly to Therefore, this F&O has been pipe damage] of RHR or SI piping is addressed. not considered in the Fire PRA.

1-8 IE-D1, F Documentation builds on The three examples cited in the Not applicable to the FPRA.

AS-C1, earlier documentation. F&O were addressed.

LE-G1, IFSN-A12, IFSN-B1 Page 57 of 136 to ULNRC-05876 F&O Associated (F)inding or No. SR(s) (S)uggestion Brief Description Disposition for IE PRA Disposition for FPRA 1-9 IE-A5 F Is loss of ESW a separate Further evaluation indicates that Finding is not applicable to the FPRA.

IE? creation of a separate Loss of Loss of ESW/SW is considered in fault ESW initiating event is not trees for FPRA.

necessary or justified.

1-141 DA-D3, F Treatment of CCF F&O was evaluated. F&O Internal Events PRA response is DA-E2 uncertainty. response includes applicable to FPRA.

recommendation to perform CCF uncertainty sensitivity runs in the future once the Data Parameter file is completed.

1-20 IFSO-A4, F Either apply applicable Resolution of this F&O is The flooding events referred to in the IFSO-B2, generic data for human- pending. This F&O is related to F&O are not used in the FPRA. [n/a]

IFEV-A7 induced flooding or the internal events IF analysis, develop plant-specific and does not impact the Fire human-induced flood PRA.

frequencies.

1-251 AS-B3 F Consideration of Resolution of this F&O is Resolution of this F&O is pending. It is phenomenological pending. It is not anticipated not anticipated that resolution of this conditions that resolution of this F&O F&O would have any impact on the would have any impact on the IE FPRA.

PRA.

2-6 LE-B1 F Probability of successful MAAP runs were performed to The FPRA Level 2 evaluation uses ex-vessel cooling should justify the probability used. probabilities and split fractions from the be justified. previous Level 2 analysis. [n/a]

Page 58 of 136 to ULNRC-05876 F&O Associated (F)inding or No. SR(s) (S)uggestion Brief Description Disposition for IE PRA Disposition for FPRA 3-1 AS-A1, F The F&O questions The need to question seal S3 tree for FPRA is not the same event AS-A2, whether the S(3) event cooling following a very small tree as for the internal events. S3 event AS-A3 tree should question loss LOCA was evaluated. It was tree for FPRA delineates very small of RCP seal cooling. determined that loss of seal LOCAs caused by fire induced events.

cooling does not need to be Loss of seal cooling caused by fire included in the S(3) event tree. induced is asked and delineated on the transient event tree in FPRA. The FPRA postulates loss of seal cooling caused by fire related events and a S3 LOCA caused by fire relate events, but does not postulate a random S3 LOCA simultaneous with a fire induced S3 LOCA. [n/a]

3-6 AS-A5, F Include potential for This F&O was evaluated. As a Residual LOSP and consequential AS-B2 consequential LOOP in result, consequential LOOP was LOSP are not included in FPRA. (A RCP seal LOCA AS added to the Tc and Tsw event LOSP resulting from a fire and/or analysis. trees. random failures inside the plant boundary is included in the FPRA.)

The assumption was reviewed and approved by the fire peer review. [n/a]

1-31 IE-C8 S Include PEG01A FTS on Calculation EG-19 Revision 1 CCF failures on LOSP are included in loss/recovery of power to was modified such that pump the FPRA. A secondary fail-to-start the pump. start failures (including common failure mode is not included for the cause start failures) for PEG01A 50% of the time pump A is running.

(the running pump) are included for loss of normal power to the running pump followed by power recovery.

1-5 IE-C9 S This F&O questions the This F&O has not been resolved. This issue pertains only to the Loss of mission time used for All CCW initiator fault CCW pressure tree/quantification. It does not impact transmitters in the CCW the FPRA.

initiator model/FT.

Page 59 of 136 to ULNRC-05876 F&O Associated (F)inding or No. SR(s) (S)uggestion Brief Description Disposition for IE PRA Disposition for FPRA 1-6 IE-C14 S Justify the 1" ISLOCA ZZ-105, Rev. 0, Add. 1 was Same justification as for Internal Events screening criterion. revised to add justification for PRA.

the 1 screening criterion (i.e.,

e) used in the ISLOCA location review.

1-12 DA-D6 S Suggested data Additional information was No fire response necessary. [n/a]

documentation added to the affected enhancements. documentation.

1-131 DA-D6 S Two potential issues The identified issues were The CCF values and basic events for identified with application addressed. the Internal Events PRA were updated of the common-cause in 2011. The work was independently data. peer reviewed in August 2011 and peer review comments were addressed and finalized in October 2011, well after the NFPA-805 LAR was submitted. Some CCF values increased and some decreased. A succinct sensitivity study for each CCF value to determine its individual effect has not been performed at this time. The effort to incorporate each of the updated CCF values in the FPRA is substantial, as it would require re-quantification of all fire scenarios and all Fire Risk Evaluations. The updated CCFs are scheduled to be incorporated as part of the next FPRA revision.

1-151 LE-C6, S The F&O suggests This F&O has not yet been This F&O is a suggestion, which is not LE-C7 consideration of pre- addressed. However, the LERF yet addressed by the Internal Events initiator CTMT isolation analysis already includes a PRA. Consequently, it has not been failures in the CTMT FAIL_LEAK event, obviating addressed by the FPRA.

isolation systems model. the need to take any action in response to this F&O.

Page 60 of 136 to ULNRC-05876 F&O Associated (F)inding or No. SR(s) (S)uggestion Brief Description Disposition for IE PRA Disposition for FPRA 1-16 IFSN-A6 S Need qualitative To address this F&O, Random flooding events are not assessment of pipe whip, documentation was added to the addressed in the FPRA.

humidity, temperature, Internal Flooding Notebook.

etc., in IF analysis.

1-18 AS-A4 S Suggestion for AS F&O response clarifies Not applicable to FPRA documentation documentation Callaways current approach to report.

enhancement. documentation.

1-19 IFSO-B2, S Flood source screening Information added to the IF Random flooding events are not IFSN-A15 documentation Notebook. addressed in the FPRA.

enhancement.

1-23 IFQU-A3, S Suggested IFQU Minor revisions were made to Random flooding events are not IFQU-B3 documentation the IF Notebook to address this addressed in the FPRA.

enhancement relative to F&O.

screening quantification decisions.

1-26 IFEV-A6 S Provide a more complete This F&O has not yet been Random flooding events are not discussion of plant- addressed in the IE PRA. addressed in the FPRA. The F&O is specific experience that related to the Internal Events PRA IF could impact flood analysis, and has no bearing on the likelihood. FPRA.

2-1 LE-A3 S Suggested LERF Additional information added to Internal Events LERF documentation is documentation LERF Notebook. not applicable to FPRA.

enhancement.

2-2 DA-A2 S Suggested addition of Additional discussion added to This is incorporated into the FPRA by component boundary, Data calculation. virtue of the fact that the random failure failure mode and success probabilities in the FPRA are from the criteria discussion to DA Internal Events PRA.

documentation.

2-3 IFPP-B1 S Documentation suggestion Minor revisions made to the IF Random flooding events are not relative to IFPP. Notebook to address this F&O. addressed in the FPRA.

2-4 LE-C1 S More justification Additional information added to Not applicable to the FPRA.

required for the definition the LERF Notebook.

used for early release.

Page 61 of 136 to ULNRC-05876 F&O Associated (F)inding or No. SR(s) (S)uggestion Brief Description Disposition for IE PRA Disposition for FPRA 2-5 LE-C5 S Suggestion relative to the Specific suggestions of the F&O F&O not applicable to FPRA LERF use of conservative versus were addressed. analysis.

realistic LERF success criteria.

2-8 IFSO-A5 S Suggestion to add Information added to IF Random flooding events are not temperature of flood Notebook in response to this addressed in the FPRA.

sources to IF F&O.

documentation.

3-2 AS-A2 S Revise ZZ-275 CSF/SC Revised tables were added to the Internal Events PRA event tree tables to match current Tc AS calculation set to address this documentation not applicable to FPRA.

and Tsw event trees. F&O.

3-5 AS-A7 S Suggestion to consider the Suggestion was evaluated, and Spurious-open PORV is a consequential need to add the potential justification for not adding a event in the FPRA, which will appear in for a stuck-open PORV to stuck open PORV to these ETs any scenario where fire damage can T1s, Tc and Tsw event was generated. cause it to occur. [n/a]

trees.

3-8 IFQU-A1, S Suggestion for additional This suggestion has not yet been Random flooding events are not IFQU-B2 documentation. addressed. However, it pertains addressed in the FPRA.

to the IFQU element, and does not impact the Fire PRA.

3-9 IFSO-A1, S Document a basis for This suggestion has not yet been Random flooding events are not IFSN-A8 floor penetrations and addressed. However, it pertains addressed in the FPRA.

block walls not failing due to the IF analysis, and does not to flood loads. impact the Fire PRA.

3-10 IFSN-A10 S Suggestion to consider the This suggestion has not yet been Random flooding events are not potential for floor drain addressed. However, it pertains addressed in the FPRA.

blockage. to the IF analysis, and does not impact the Fire PRA.

4-1 LE-C1 S Suggested minor revision Minor revision made to LERF F&O is not applicable to the FPRA.

to LERF-related text. Notebook to address this F&O.

Page 62 of 136 to ULNRC-05876 F&O Associated (F)inding or No. SR(s) (S)uggestion Brief Description Disposition for IE PRA Disposition for FPRA 4-2 LE-C8 S Suggestion for additional Information added to LERF F&O is not applicable to the FPRA.

discussion of how Level 1 Notebook.

and Level 2 models are linked.

4-3 AS-A7, S Suggested enhancement Additional text justification was The event tree assumptions that are AS-A10 of RCP seal LOCA developed, and will be included critiqued here are not made in the accident sequence in an AS Notebook, currently FPRA. F&O does not apply to the documentation. under development. FPRA.

4-4 LE-G2 S Suggestion relative to Information added to LERF F&O is not applicable to the FPRA.

LERF documentation Notebook.

enhancement.

4-5 LE-C9 S Suggestion to justify Justification/text added to the F&O is not applicable to the FPRA feedwater availability LERF Notebook. because the status of feedwater after core damage (as availability at the time of core damage credited in the LERF or after core damage is not considered model). in the FPRA LERF model.

Note 1 - These F&Os will be evaluated and/or incorporated into the fire PRA during the next update, per Implementation Item 12-805-001 shown in the updated Attachment S to this enclosure.

Page 63 of 136 to ULNRC-05876 Probabilistic Risk Assessment RAI 02 The peer review description addresses the relevant internal events PRA standard, but does not identify how the Regulatory Guide (RG) 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," March 2009 (ADAMS Accession No. ML090410014), clarifications and qualifications to the standard were addressed. Please indicate whether or not RG 1.200 clarifications and qualifications to the standard were considered by the peer review team, and, if not, provide a self-assessment of the PRA model for the RG 1.200 clarifications and qualifications and indicate how any identified gaps were dispositioned.

This also applies to the FPRA peer review. In your response, please address both peer reviews.

Response to PRA RAI-02 Response provided by ULNRC-05851 dated April 17, 2012.

Page 64 of 136 to ULNRC-05876 Probabilistic Risk Assessment RAI 03 The disposition of the F&Os related to Large Early Release Frequency (LERF) refers to a separate LERF model developed for the FPRA. Please provide a discussion of the peer review of this new LERF model and identify and disposition any peer review F&Os.

Additional Justification Needed The response seems to say that a copy of the LERF event tree was used only to simplify linking between the trees. However, in the response to an audit question, it is explained that a simplified version of the LERF event tree was used, what was simplified, and why this is OK. This more descriptive response should be included to the actual RAI 3 response.

Response to Probabilistic Risk Assessment RAI-03 The Callaway Plant FPRA developed a LERF model specifically for the FPRA from the LERF model for the Internal Events PRA. The development was necessary to enable the FPRA LERF model to run from the same set of event trees as the Level 1 fire sequence development and run from the same PRA batch file as the Level 1 FPRA core damage sequences.

This model was complete and in place at the time of the FPRA peer review in October 2009. Although there are no specific Fire Supporting Requirements for LERF, deficiencies in the LERF model would be identified from the following Fire SRs, which relate back to Internal Events SRs from Chapter 2 of the combined PRA Standard:

PRM-B1 PRM-B14 PRM-B15 PRM-C1 FQ-D1 FQ-E1 FQ-F1 UNC-A2 A review of the F&Os from the Callaway Plant FPRA peer review shows 2 F&Os which cited LERF Findings. These are shown in the table below, along with their resolution.

Page 65 of 136 to ULNRC-05876 DESCRIPTION OF F&O SUGGESTION SR RELATED TO LERF OR FINDING RESOLUTION PRM -C1 Documentation needs to be expanded. The Finding Documentation was following are examples of documentation expanded in section that should be provided to support the Plant 4.8.1 of the fire-Response Model (PRM) and quantification: induced risk model report (17671-004) to (8) Document the linkage of the containment address this F&O.

isolation in the LERF calculation.

FQ -F1 Documentation was limited for fire Finding Documentation was quantification. expanded in sections 5.2 and 5.3 of the (5) Document the significant contributors to integrated fire risk the Fire PRA. Document both the dominant report (17671-013) to LERF and CDF contributors. address this F&O.

The FPRA developed a LERF model based on the 2006 version of the Callaway Plant Internal Events PRA. The 2006 Internal Events PRA LERF analysis develops a LERF equation which considers a) core-containment energetics, b) ISLOCA, c) steam generator tube rupture, d) containment isolation failure and leakage. The Internal Events PRA uses a containment event tree to develop a LERF split fraction representing LERF due to containment failure from core - containment energetics. Items (b),

(c), and (d) from above are modeled explicitly.

The FPRA used the same process for LERF model development, except that LERF split fraction due to core-containment energetics for part (a) above are derived directly from the Internal Events PRA, based on similarity of Plant Damage States (PDS). Potential PDSs for the fire sequence were compared to the PDSs from the Internal Events PRA sequences and the maximum LERF split fraction for any PDS resulting from any applicable Internal Events PRA PDS was used in the FPRA. Items (b), (c), and (d) from above were explicitly calculated for the FPRA given the applicable specific failures in the fire scenario.

The Internal Events PRA was recently updated. The LERF model for the Internal Events PRA was updated and finalized in December 2011, after the NFPA 805-license amendment request was submitted to the NRC. The FPRA LERF model will be updated to be consistent with the Internal Events PRA LERF model phenomenology, split fractions, and probabilities during the next FPRA update per Implementation Item 12-805-003. Implementation Item 12-805-003 is shown in the updated Attachment S to this enclosure.

Overall, the LERF / CDF split fraction in the current Internal Events PRA model is 1.3% (i.e., 2.29E-5 CDF / 3.09E-7 LERF) versus a LERF ratio of 1.97% for the FPRA (2.03E-5 CDF / 3.99E-7 LERF).

In accordance with RG 1.205 and RG 1.174, LERF does not become the determining metric for an application unless the LERF / CDF split fraction it is greater than 10%. The LERF fractions for the Callaway Plant FPRA and the Internal Events PRA are similar and are both significantly less than 10%.

Page 66 of 136 to ULNRC-05876 Probabilistic Risk Assessment RAI 04 The resolution of a number of F&Os from the internal events review does not appear to fully address the impact of the resolution on the NFPA-805 results. Please justify the proposed resolution as follows:

a. During the audit, the licensee identified that F&O SY-2 disposition should be revised to indicate that the item was addressed in the FPRA. Please provide this revised disposition.

b. F&O DA-3 discusses basic events and sensitivity studies conducted. The licensee identified during the audit that this item was in fact resolved for the FPRA. Please provide this revised disposition.

c. F&O IE-8 (which is cross-referenced to Supporting Requirement DA-C14) identifies recovery events in the internal events model which may not have appropriate probabilities.

The disposition of this F&O states that the FPRA does not "generally" credit these recovery actions. Please provide a more substantive justification that this F&O is not relevant to the FPRA.

d. F&O IE-13 relates to the age of the Inter-System Loss of Coolant Accident (ISLOCA) evaluation (and it is assumed that changes may be needed when it is updated). During the audit, the licensee identified that a revised ISLOCA evaluation was created for the FPRA and was peer reviewed. Please provide a statement indicating this and provide the disposition of any F&Os from the peer review. The FPRA peer review would not normally review the ISLOCA. Please provide a statement indicating that this was specifically included in the scope of the peer review for the FPRA.

e. F&O IF-D5/D5a relates to internal flooding gaps. Please indicate if there are any fire-induced floods (i.e., due to spurious valve opening).

f. The licensee reported (via 10 CFR 50.72) the use of high density polyethylene piping in the Essential Service Water (ESW) system that was not protected by a fire barrier in fire areas C-1, D-1, and D-2, but Attachment W indicates no Variance from Deterministic Requirements (VFDR) in area C-1. Please provide a discussion on how this design deficiency has been addressed and provide any required changes to the NFPA 805 LAR.

This should include, as appropriate, FPRA modeling considerations, VFDR identification, and a discussion of the fire scenarios which challenge the integrity of the piping (i.e., HRR levels assumed for transient combustible ignition sources in the analyses performed to address the use of this polyethylene piping.)

Additional Justification Needed:

No Storage and No Hot Work are cited as the basis for assuming a transient combustible HRR of 69 kW (98th %ile) in Area C-1, containing high-density polyethylene piping. In response to PRA RAI 23, multiple reasons were cited as the basis for the lower Page 67 of 136 to ULNRC-05876 HRR assumption (69 kW). Which of these other factors are also applicable in Area C-1, (i.e., not just the designations of No Storage and No Hot Work). Also, the response cites changes made to the Transition Report (LAR) reflecting an updated analysis for Area C-1, specifically Table 4-3, Att. C, Att. D and Table W-2. (Submit LAR update)

Response to Probabilistic Risk Assessment RAI-04

a. Response provided by ULNRC-05851 dated April 17, 2012.

b. Response provided by ULNRC-05851 dated April 17, 2012.

c. Response provided by ULNRC-05851 dated April 17, 2012.

d. Response provided by ULNRC-05851 dated April 17, 2012.

e. Response provided by ULNRC-05851 dated April 17, 2012.

f. Background Callaway Plant reported as a Licensee Event Report (LER), a condition identified in the plant involving High Density Polyethylene (HDPE) piping that affects the NFPA 805 License Amendment Request (LAR) and its supporting documentation (reference LER 2011-006-00 transmitted via ULNRC-05836 dated 1/6/12). During a review of the analysis associated with Fire Area C-1, Pipe Space and Tank Area, Control Building Elev. 1974 / 1984 it was determined that the HDPE piping that had been installed by plant modification could be affected by a fire. The resulting HDPE pipe failure could create a flooding condition where one train of required Essential Service Water (ESW) equipment is not maintained free of fire damage. Fire Area C-1 (CB 1974' room 3101) contains both trains of ESW supply and return piping. Fire Protection Program compliance in Fire Area C-1 is met by the ESW trains being separated by 20 feet with no intervening combustibles and automatic detection and suppression.

NFPA 805 LAR Impacts HDPE Piping During development of the NFPA 805 LAR and the associated Nuclear Safety Capability Assessment (NSCA) for fire area C-1, the impact of a fire on the HDPE pipe was not considered. Fire Area C-1 is comprised of two fire zones (rooms), room 3104 which is a stairwell and room 3101 which is the large pipe space that contains the ESW system related equipment. Fire Area C-1 was determined to be deterministically compliant based on the two trains of ESW piping and valves to be adequately separated by 20 feet of separation with the presence of automatic detection and suppression in Room 3101. Additionally, fire modeling was performed in Room 3101 and as a result of FPRA risk insights, fire area C-1 Room 3101 had been designated a No Storage Location and No Hot Work Location. The only ignition sources in Room 3101 are transient hot work and transient combustibles. To address the fire Page 68 of 136 to ULNRC-05876 induced HDPE pipe failure scenario, two new variances from the deterministic requirements (VFDRs) were developed for Fire Area C-1; one for each ESW train of HDPE piping in Room 3101.

A Fire Risk Evaluation (FRE) was completed for Fire Area C-1 to evaluate the impact of the new VFDRs and the results are documented in a revised C-1 Fire Safety Analysis (FSA),

calculation KC- 113, Fire Safety Analysis for Fire Area C-1. To support the FRE, fire modeling was conducted in fire area C-1 using Fire Dynamics Simulator (FDS) to evaluate the impact of transient fire scenarios in the immediate vicinity of the HDPE pipe for their ability to damage the pipe. Because Fire Area C-1 is considered a No Storage and No Hot Work area, the HRR postulated for the transient fires is 69 kW. C-1 is a pipe chase with limited equipment and large combustible liquid fires are not expected. Since only small quantities of trash in temporary containers can be expected, a 69kW 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. The FDS fire modeling demonstrated that the HDPE pipe will remain free of fire damage. The FRE also evaluated defense in depth and safety margin and determined a Main Control Room operator action was required to align ESW valves which ensures that they are in the required NSCA position should flooding occur.

A new delta risk calculation was developed for Fire Area C-1, calculation 17671-FRE-C-1, Fire Risk Evaluation for Fire Area C-1. Because the VFDRs (pipe) are not damaged by any fire scenario damage set, the delta risk for the VFDRs is negligible and overall absolute fire risk in Room 3101 remains unchanged.

The Callaway Plant Transition Report has been revised to change the NFPA 805 regulatory basis for fire area C-1 from 4.2.3.2 to 4.2.4.2 in Transition Report Table 4-3. In addition, Transition Report Attachments C, D and W, have been revised. The VFDRs have been added to the Fire Area C-1 discussion in Attachment C, NEI 04-02 Table B-3 Fire Area Transition, Attachment D Non Power Operational Modes Transition and Attachment W, "Fire PRA Insights". These changes are reflected in Attachments 1, C, D and W of this enclosure.

Additionally, background documents have been updated to reflect the VFDRs and other required description changes as follows;

• Calculation KC-113, Fire Safety Analysis for Fire Area C Revised to include a description of HDPE pipe, to add the VFDRs, and revise the fire area boundary description to include the prefabricated enclosure used at the ESW pipe wall penetration.

• Calculation KC-116, Fire Safety Analysis for Fire Area C Revised to reflect the change to the description of the fire barrier interface with C-1.

• Calculation KC-117, Fire Safety Analysis for Fire Area C Revised to reflect the change to the description of the fire barrier interface with C-1.

Calculation KC-57, Detailed Fire Modeling Report for Fire Area C Revised to include a description of HDPE pipe, to revise the fire area boundary description to include Page 69 of 136 to ULNRC-05876 the prefabricated enclosure used at the ESW pipe wall penetration and to include a discussion of the FDS fire modeling performed.

• Calculation KC-58, Detailed Fire Modeling Report for Fire Area C Revised to reflect the change to the description of the fire barrier interface with C-1.

• Calculation KC-59, Detailed Fire Modeling Report for Fire Area C Revised to reflect the change to the description of the fire barrier interface with C-1.

• Calculation R1984-001-001, Fire Dynamics Simulator (FDS) Analysis To Support Detailed Fire Modeling - Revised to add the FDS fire modeling which evaluated transient fire impacts to the HDPE pipe.

• Calculation KC-26, Nuclear Safety Capability Assessment - Revised to address the VFDRs and their resolution.

• Fire Pre-Plan Manual- Revised fire response guidance for C-1 to identify HDPE piping locations.

• Calculation 17671-FRE-C-1, Fire Risk Evaluation for Fire Area C New document Elastomeric Isolation Joints During an extent of condition evaluation for the HDPE pipe issue described above, additional elastomeric components were identified in the plant that if failed by fire could have a potential for adverse consequences on protected train equipment. In Fire Areas D-1 and D-2, the two emergency diesel generator (EDG) rooms, there is ESW piping which contains elastomeric expansion joints. Should an elastomeric joint fail due to fire exposure the resultant flooding in the fire area could impact adjacent fire areas and affect redundant train equipment if no operator response is taken.

Each EDG is cooled by ESW, which has both supply and return piping containing elastomeric expansion joints, that if failed due to fire damage will result in ESW flooding the EDG room.

For Fire Area D-1 if no action is initiated to isolate the ESW water flow, the resultant flooding could impact the adjacent fire area which contains the credited train of electrical switchgear.

During development of the NFPA 805 LAR and the associated NSCA analysis, Fire Areas D-1 and D-2 were determined to be deterministically compliant with the requirements of NFPA 805 Chapter 4. The elastomeric components failures were evaluated and determined not to meet the criteria of a VFDR as the failed condition is recovered by a Main Control Room action. The Callaway Plant LAR is not affected; however, LAR background documents have been revised as follows; Calculation KC-149, Fire Safety Analysis for Fire Area D Revised to include a description of the expansion joints and the impact.

Page 70 of 136 to ULNRC-05876 Calculation KC-150, Fire Safety Analysis for Fire Area D Revised to include a description of the expansion joints and the impact.

Calculation KC-75, Detailed Fire Modeling Report for Fire Area D Revised to reflect the change to the damage sets for fire scenarios that are in the vicinity of the elastomeric joints.

Calculation KC-76, Detailed Fire Modeling Report for Fire Area D Revised to reflect the change to the damage sets for fire scenarios that are in the vicinity of the elastomeric joints.

Calculation KC-26, Nuclear Safety Capability Assessment - Revised to address a main control room action to isolate ESW.

Fire Pre-Plan Manual- Revised fire response guidance for fire area D-1 and D-2 to identify that elastomeric joints may create flooding conditions.

OTO-KC-00001, Fire Response, Addendums D-1 and D Revised to add a Main Control Room step to secure the affected train ESW pumps to mitigate the flooding.

Page 71 of 136 to ULNRC-05876 Probabilistic Risk Assessment RAI 05 If changes to the FPRA model have been made subsequent to the completion of the peer review of the FPRA, please provide a description of any new models or methods that have been implemented for the FPRA, including any subsequent focused scope peer reviews of the models or methods.

Response to Probabilistic Risk Assessment RAI-05 Response provided by ULNRC-05851 dated April 17, 2012.

Probabilistic Risk Assessment RAI 06 The resolution of a number of F&Os from the FPRA review does not appear to fully address the impact of the resolution on the NFPA-805 results. Please justify the proposed resolution as follows:

a. F&O ES-A1 This F&O disposition identifies an "updated generic list of multiple spurious operations (MSOs)" to be considered to resolve this item. The disposition does not explicitly state the updated list was used, only that the "generic pressurized water reactor (PWR) MSO list" was reviewed. Please clarify this response.

b. F&O ES-B1 It is not clear what the deficiency in the FPRA model is, or if the item was resolved by making changes or by simply clarifying the underlying issue. Please clarify this and discuss how it was addressed.

c. F&O ES-B2 Flow diversion paths screened in the internal events PRA due to low frequency may become significant due to spurious operations. Please provide a description of the method for consideration of diversion pathways which could be significant in the FPRA model due to a spurious operations failure mode.

d. F&O ES-C1 The disposition is not clear as to whether a change was made to address the F&O, or if it is providing the location of the missing information which was simply not found by the peer review team (i.e., it is not a valid F&O). Please provide clarification as to how the F&O was addressed.

e. F&O CS-B1 The disposition is not clear as to whether a change was made to address the F&O, or if it is providing the location of the missing information which was simply not found by the peer review team (i.e., it is not a valid F&O). Please provide clarification as to how the F&O was addressed.

f. F&O FSS-B01 The F&O has two distinct parts. The first part is partially addressed by the evaluation of a specific cabinet in the control room which can cause a loss of heating ventilation and air conditioning (HVAC), which stated that an updated analysis considers a fire spreading to this cabinet, but the response does not specifically address a fire originating in the cabinet. The second part, the potential complexity of a fire event causing Page 72 of 136 to ULNRC-05876 spurious safety injection (SI) and containment isolation, is not addressed in the disposition.

Please provide a more complete disposition of this F&O.

Response to Probabilistic Risk Assessment RAI-06 Response provided by ULNRC-05851 dated April 17, 2012.

Page 73 of 136 to ULNRC-05876 Probabilistic Risk Assessment RAI 07 A number of issues with the Human Reliability Analysis (HRA) need to be clarified:

a. F&O HRA-E1 The F&O indicates that the human error analysis credits instrumentation not traced to assure availability. Please confirm that credited instrumentation relied upon for the HRA is based on availability of instruments free from fire damage.

Additional Justification Needed:

While it likely may be inferred, the RAI requested confirmation that available instruments were free from fire damage. The response cites the list of available instruments but does not explicitly state that they are free from fire damage when being credited. Provide this confirmation if correct. If not, explain the basis for taking credit.

b. Please provide the basis for assuming a screening human error probability of 0.1 for failure of successful operation at the Auxiliary Shutdown Panel (ASP) following Main Control Room (MCR) abandonment.

c. Some of the time windows cited in the Post-Fire HRA Calculation to complete a task seem very short (e.g., (1) HFE OP-OMA-FF-EGRVAB (probability = 0.13) with a "time margin" of only one out of 20 minutes (min), where the available time (20 min) is an assumption based on a hand calculation; (2) HFE OP-OMA-FF-ISOEG (probability = 0.5), with no "time margin" and an available time based on a conservative hand calculation; (3) HFE OP-OMA-FF-RCPTRP (probability = 0.29), with a "time margin" of only 0.8 out of 13 min.

Note that, for this third example, ranges on various time frames based on discussions with plant personnel are cited when estimating the total time to execute (~9 min). If the upper ends of the cited ranges are assumed, this execution time becomes ~10.5 min which, when combined with the assumed 5-min delay time, exceeds the available time by ~1.5 min.

Please discuss whether or not (1) the methodology was reviewed in the peer review and (2) the methodology was consistently applied to all HRA. Please include the results of a sensitivity evaluation if each human error probability is assumed to be 1.0 (or some other bounding value, with justification), or provide the basis for the assumed value being appropriate.

Additional Justification Needed Do the calculational results (Tables 2 through 4) reflect removal of credit for CPTs in all scenarios? If not, how would the results change if this credit were removed entirely? Note that, for the delta calculations, the credit should be removed in both the base and comparative cases.

d. F&O FQ-C1 The F&O identifies that the HRA dependency analysis does not consider execution dependencies for local actions for fire scenarios. This item is indicated to be closed, but the disposition is to review and disposition these dependencies in the next Page 74 of 136 to ULNRC-05876 FPRA update. Please confirm completion of this item sufficient to resolve the technical issue for the existing fire PRA used to support this application.

e. Conservatism in the current state of FPRA was cited as the basis for: (1) considering it premature to perform a detailed dependency analysis for the fire HRA; (2) dismissing completeness uncertainty as a current concern in fire HRA; (3) not performing uncertainty analysis on fire risk and delta-risk. Please provide either: (1) sensitivity evaluations to address the potential impact of not explicitly addressing these issues or (2) a discussion of the plant-specific aspects of the FPRA for Callaway that constitute the basis for the cited conservatism.

Additional Justification Needed Table 3-1 of Calculation 17671-014 is cited as providing discussion as to why completeness uncertainty does not apply to fire HRA. State explicitly this material from Table 3-1. In addition, Table 4-3, presumably of the same calculation, is cited as providing importances of applicable recovery actions. However, there appears to be no such table in the calculation. If Table 4-3 of the LAR is meant, note that this does not address HRA, but rather Fire Protection Systems and Features. Provide clarification and, if necessary, correction.

f. If a sensitivity/uncertainty analysis was performed for the Fire LERF and Delta-LERF (LERF) after the LERF model was ready, please report the results. If not, please perform an analysis or justify the basis for assuring that the insights to be gained from a sensitivity/uncertainty analysis were obtained otherwise and the means of doing so.

Additional Justification Needed "Report the results" means that they should actually be docketed, either in this RAI response or as part of an update to the LAR, not just referenced as available in a portal document (17671-014, App. B). This includes the results for CDF, LERF, delta-CDF and delta-LERF. Provide, e.g., one of the following: (1) add the material from the portal document to the RAI response, or (2) embed this material in the LAR as updated. Keep in mind the potential effect of the response to RAI 9b on the material in 17671-014, App. B (Sensitivity #1).

Response to Probabilistic Risk Assessment RAI-07 a) F&O HRA-E1-1 (Suggestion level F&O) was written in October 2009 during the Fire Peer Review. Since 2009, the process described below was implemented to ensure the human reliability analysis only credited instruments that are free of fire damage.

To start, Table 3-3 of the Callaway Plant Fire HRA report (Calculation 17671-011) provides the specific instruments that are required by operator actions in order to accomplish diagnosis.

Each of the instruments listed in Table 3-3 was cable traced. Table 3-3 was updated following the 2009 FPRA peer review to include references to specific instruments.

Page 75 of 136 to ULNRC-05876 Next, in order to ensure the availability of instrumentation for operator actions, human failure events (HFEs) were grouped into two categories.

1) HFEs directed by the fire response procedures and evaluated as part of the NSCA (Calculation KC-26). These HFEs are shown in Table 3-4 (Local Actions) and 3-5 (Control Room Actions) of the Callaway Plant Fire HRA report (Calculation 17671-011). Part of the NSCA evaluation of human actions is to ensure adequate instrumentation and cues are available to support the action on an area-by-area basis wherever the action is required. For these HFEs, the PRA used the NSCA analysis to ensure that instrumentation and procedures are available to support the action.

2) HFEs directed by the internal events Emergency Operating Procedures (EOPs). Table 3-3 of the Callaway Plant Fire HRA report (Calculation 17671-011) lists the HFEs associated with this category and provides the specific instruments that are required for diagnosis. Each of the instruments listed in Table 3-3 was cable traced.

Instrument Availability Considerations During Quantification:

Two trains of instrumentation are typically provided for any operation. Human error probabilities (HEP) values are developed for three cases of instrumentation availability:

a) Both trains of instrumentation available. HEP at a nominal value that accounts for fire-effects.

b) One train of instruments available. HEP increased in accordance with guidance in NUREG-1921.

c) No instrumentation available. HEP is guaranteed failed.

In order to more efficiently conduct the fire scenario quantification while providing a conservative basis for the HEPs, HEPs were calculated as case (b) crediting only a single train of instrumentation. This is a higher HEP value than the nominal fire value, but precludes the needs to match a specific HEP with every fire scenario. For all instruments in Table 3-3, it was verified that at least one train of instruments were available for each fire area, so that condition (b) applied. If this was not the case, the HEP was assigned a value of 1.0.

b) Response provided by ULNRC-05851 dated April 17, 2012.

c) The human reliability analysis (HRA) methodology used to calculate the human error probabilities (HEPs) for the three cited events was consistent with the method that was applied to all human failure events (HFEs) in the Callaway Plant FPRA. This method was examined during the October 2009 Peer Review and there are no HRA-related peer review comments that are open.

There are three HFEs involved in this question. These HFEs are shown in Table 1 below, along with the Fire Areas in which they are credited, and their nominal HEPs.

Page 76 of 136 to ULNRC-05876 Table 1 - Applicable Human Failure Events Credited Fire Nominal Event Description Areas Probability OP-OMA-FF-EGRVAB Operator fails to locally close spuriously C21 / C22 / C27 1.3E-1 open CCW surge tank vent valve for CCW train swap.

OP-OMA-FF-ISOEG Operator fails to locally close spurious A21 / C21 / 5.0E-1 open CCW/ESW fill valves. C22 / C24 /C27 OP-OMA-FO-RCPTRP Operator fails to locally trip the RCPs C27 2.9E-1 when MCR trip capability is failed.

A sensitivity study was performed to find the CDF/LERF/delta CDF/delta LERF changes if no credit is allowed for the HEPs in Table 1. This is a bounding case.

In addition, in order to combine sensitivity studies, the changes were done for the case with and without credit for the control power transformer (CPT) in determining the probabilities for spurious valve actuation (issue from RAI-PRA-9b). The existing License Amendment Request (LAR) results, which credit these HEPs at their nominal values, and the sensitivity results are compared in Tables 2a and 2b below.

Note that all results tables have an a and a b version. The a versions show risk results that credit control power transformers (CPT) during the calculation of spurious actuation probabilities based on fire-induced cable damage. The b version of each table show risk results with the CPT credit removed.

Table 2a - LAR vs. HEP Sensitivity Results (with CPT credit)

LAR Results HEP = 1.0 Sensitivity Results CDF LERF CDF LERF CDF LERF CDF LERF (per yr.) (per yr.) (per yr.) (per yr.) (per yr.) (per yr.) (per yr.) (per yr.)

8.07E-06 2.50E-07 3.87E-06 2.03E-07 8.45E-06 2.86E-07 4.22E-06 2.38E-07 Table 2b - LAR vs. HEP Sensitivity Results (without CPT credit)

LAR Results HEP = 1.0 Sensitivity Results CDF LERF CDF LERF CDF LERF CDF LERF (per yr.) (per yr.) (per yr.) (per yr.) (per yr.) (per yr.) (per yr.) (per yr.)

2.29E-05 5.04E-07 6.24E-06 2.77E-07 2.35E-05 5.41E-07 6.78E-06 3.14E-07 Page 77 of 136 to ULNRC-05876 The increase in each metric is shown in Tables 3a and 3b.

Table 3a - Risk Metric Increases (with CPT credit)

CDF LERF CDF LERF (per yr.) (per yr.) (per yr.) (per yr.)

3.79E-07 3.55E-08 3.49E-07 3.48E-08 Table 3b - Risk Metric Increases (without CPT credit)

CDF LERF CDF LERF (per yr.) (per yr.) (per yr.) (per yr.)

5.80E-07 3.79E-08 5.41E-07 3.70E-08 The increases in CDF and LERF are compared to the plant totals in Tables 4a and 4b.

Table 4a - Delta Risk Increases Relative to Plant Totals and Regulatory Limits (with CPT)

RAI 07c Metric LAR Increase Total Limit CDF (per yr.) 1.87E-06 3.49E-07 2.22E-06 < 1E-5 LERF (per yr.) 3.84E-08 3.48E-08 7.32E-08 < 1E-6 Table 4b - Delta Risk Increases Relative to Plant Totals and Regulatory Limits (without CPT)

RAI 07c Metric LAR Increase Total Limit CDF (per yr.) 1.87E-06 5.41E-07 2.41E-06 < 1E-5 LERF (per yr.) 3.84E-08 3.70E-08 7.54E-08 < 1E-6 This calculation has taken the HEPs in question and bounded their uncertainty by setting them to 1.0 in the sensitivity quantification. Although the HRA methods used to calculate the nominal HEP values were reviewed during the FPRA Peer Review and are consistent with those of the rest of the Fire PRA, this analysis has been performed to show the worst-case scenario.

As shown in Tables 4a and 4b, even with this bounding approach, if the delta risk impact of this sensitivity study is added to the existing LAR delta risk totals, the resulting totals are still comfortably below the regulatory limits. Additionally, the final risk insights shown in Tables 4a and 4b indicate that the conclusions of this sensitivity study are not significantly impacted by removing credit for the CPTs in the spurious actuation probability calculations. The results without CPT credit in Table 4b show a slightly higher increase in delta risk compared to the CPT-credited results in Table 4a, but both sets of results maintain acceptable margin to the regulatory limits.

Page 78 of 136 to ULNRC-05876 The nominal HEP values used for the LAR risk calculations were calculated with the standard FPRA HRA methods used throughout the model, which were reviewed during the Peer Review in October 2009. However, even if the HEPs are set to 1.0 to bound potential uncertainty in the HEP values, the overall risk insights are unaffected.

d) Response provided by ULNRC-05851 dated April 17, 2012.

e) The wording of the Callaway Plant fire HRA report could have been better structured to convey the detailed dependency analysis conducted as part of the fire HRA. The Callaway Plant Fire HRA conducted a detailed dependency analysis as described in the response to PRA RAI 7d. The peer review observation of the Callaway Plant HRA dependency analysis was a Suggestion level F&O to conduct a review and ensure that combinations of execution HFEs that occur in the FPRA cutsets are actually independent. A review of the cutsets was conducted (by visual inspection) during the Fire Risk Evaluation process that examined each risk-significant fire area. Since the Callaway Plant Fire HRA accounts for dependencies in two ways and the results of the visual inspection did not find any additional dependencies, this Suggestion level F&O is considered closed. However, since the fire response procedures are being updated and trained upon as part of the transition to NFPA 805, it is recognized that the Fire HRA dependency will need to be re-visited during the implementation phase as part of Implementation Item 11-805-090 of Table S-3 in the Callaway Plant NFPA 805 License Amendment Request Attachment S.

Completeness uncertainty does apply to fire HRA, as described in the text of Section 3.3.1 of the Uncertainty and Sensitivity Analysis (Calculation 17671-014) and implemented in Table 3-

1. Table 3-1 documents sources of uncertainty in the Callaway Plant FPRA. Columns 1 through 3 list each NUREG/CR- 6850 FPRA task and summarize the generic treatment of uncertainty issues based on Appendix V of NUREG/CR-6850. Appendix V segregates these issues as either relating to uncertainty, or relating to accuracy and completeness. Since the Fire HRA task uses other NUREG/CR-6850 tasks such as component selection, plant response model, and fire modeling as inputs then the completeness of these inputs affects the completeness of the Fire HRA.

Additionally during the Fire Risk Evaluation process uncertainty was considered as described in Section 3.7 and as documented in Table 4-3 and Table 4-5 of each Fire Risk Evaluation report (Calculation 17671-FRE-X-YY, where X-YY is the individual fire area analyzed). Table 4-3 documented the importance of applicable recovery actions by showing the risk increase if the recovery action failure probability was increased. As a future task during the implementation phase, the treatment of uncertainty related to the Fire HRA will be updated as part of implementation item number 11-805-090 of Table S-3 in the Callaway Plant NFPA 805 License Amendment Request.

f) The LERF model and the core damage model were developed in parallel and have both been available for sensitivity studies since the October 2009 peer review. All sensitivity studies report risk results for CDF, LERF, CDF, and LERF.

Sensitivity studies were performed for two considerations.

Page 79 of 136 to ULNRC-05876

1) Sensitivity studies were performed during the Fire Risk Evaluations to determine effectiveness of a recovery action. These are documented in the respective FRE reports.

2) Sensitivities were done on the global core damage and LERF equations as part of Task 14.

The Callaway Plant Fire PRA reported two sensitivity studies, which are documented in Appendix B of Calculation 17671-014, Uncertainty and Sensitivity Analyses. One sensitivity study was the FAQ 08-0048 requirement to use NUREG/CR-6850 ignition frequencies for ignition bins which are described by a gamma distribution which has an alpha factor of less than 1.0. The other sensitivity study considered the potential risk impact of small quantities of thermoplastic cable, which were discovered in limited locations in the plant. Both of these sensitivity studies considered core damage and LERF metrics, the results of which are reproduced below.

There are two sensitivity studies discussed in this RAI response: the FAQ 08-0048 ignition frequency study and the thermoplastic cable sensitivity. The FAQ 08-0048 sensitivity study results are shown in Table 1, which is a copy of Table B.1-2 of report 17671-014. This shows the total for each risk metric for fire scenarios that have ignition sources from the applicable bins, AND that contribute a non-zero delta risk to the plant-wide results.

The thermoplastic sensitivity looked at two different effects of the non-IEEE-383 cable. The first effect was the reduction of damage threshold temperature for the cables as fire targets.

Essentially, this effect increased the target set for certain fires because the non-IEEE-383 cable fails at a lower temperature, so the ZOI with respect to those cables is larger. The effect of modifying the ZOI to account for these cables is shown in Tables 2 and 3, which correspond to Tables B.2-1 and B.2-2, respectively, in the 17671-014 report.

As shown in Table 2, there was no change in risk for the affected fixed ignition sources. The affected transient sources showed a small risk increase. The last portion of the non-IEEE-383 cable sensitivity assigned the full self-ignited cable bin (bin 12) ignition frequency to each of the affected fire scenarios individually, and the scenario with the largest risk increases was selected as the bounding example. Since the full bin 12 ignition frequency was added to each scenario individually, the results should be looked at independently for each scenario, not summed. The results are shown in Table 4, with the largest risk increase being attributed to scenarios 3402-T1 and 3502-T2.

As shown in Table 4, the bounding risk increase due to the potential for self-ignited cable fires in non-IEEE-383 rated cable is 8.78E-7/yr. for fire CDF and 2.28E-8/yr. for fire LERF. There is no potential increase in delta risk because the VFDRs in the Turbine Building are not modeled in the fire PRA. These risk metric increases are very conservative because the entire bin 12 frequency was applied to each single fire scenario. A more realistic treatment would split the bin 12 frequency amongst the various fire scenarios based on cable weighting factors.

Page 80 of 136 to ULNRC-05876 Table 1: FAQ 08-0048 IEF Sensitivity Results Summary

[B.1-2 in 17671-014]

Risk Metric Baseline IEF Adjusted Delta Plant Totals (EPRI) (NUREG 6850) IEF (Baseline)

CDF (/yr) 3.17E-06 6.81E-06 3.64E-06 2.02E-05 LERF (/yr) 6.39E-08 1.42E-07 7.77E-08 3.97E-07 CDF (/yr) 1.31E-06 3.26E-06 1.95E-06 1.87E-06 LERF (/yr) 2.78E-08 7.28E-08 4.50E-08 3.84E-08 Page 81 of 136

Enclosure 1 to ULNRC-05876 Table 2: Fixed Sources with Modified ZOI

[B.2-1 in 17671-014]

Affected Final Baseline Thermoplastic Delta Ign Fire IEF CDF LERF CDF LERF CDF LERF Source Scenario (/yr) CCDP (/yr) CLERP (/yr) CCDP (/yr.) CLERP (/yr.) (/yr.) (/yr.)

4316-1 HF187A 2.79E-05 6.14E-04 1.71E-08 1.49E-05 4.16E-10 6.14E-04 1.71E-08 1.49E-05 4.16E-10 0.00E+00 0.00E+00 4316-2 HF187B 2.79E-05 6.14E-04 1.71E-08 1.49E-05 4.16E-10 6.14E-04 1.71E-08 1.49E-05 4.16E-10 0.00E+00 0.00E+00 Table 3: Transient Sources with Modified ZOI

[B.2-2 in 17671-014]

Affected Base Baseline Thermoplastic Baseline Thermoplastic Delta Fire IEF Wg*SF Final Wg*SF Final CCDP CLERP CDF LERF CDF LERF CDF LERF Scenario (/yr.) *Pns IEF *Pns IEF (/yr.) (/yr.) (/yr.) (/yr.) (/yr.) (/yr.)

4316-T1 3.25E-03 1.44E-03 4.67E-06 1.95E-03 6.34E-06 6.15E-04 1.49E-05 2.87E-09 6.97E-11 3.90E-09 9.46E-11 1.03E-09 2.49E-11 4316-T3 3.25E-03 1.17E-01 3.79E-04 1.16E-01 3.77E-04 4.49E-04 9.93E-06 1.70E-07 3.76E-09 1.70E-07 3.75E-09 -8.55E-10 -1.89E-11 3402-T1 2.44E-04 2.33E-02 5.67E-06 6.46E-02 1.58E-05 6.65E-04 1.73E-05 3.77E-09 9.79E-11 1.05E-08 2.72E-10 6.71E-09 1.74E-10 3402-T2 2.44E-04 9.77E-01 2.38E-04 9.35E-01 2.28E-04 6.15E-04 1.49E-05 1.46E-07 3.55E-09 1.40E-07 3.40E-09 -6.26E-09 -1.52E-10 3502-T2 3.56E-04 2.18E-02 7.78E-06 3.40E-02 1.21E-05 6.65E-04 1.73E-05 5.18E-09 1.34E-10 8.06E-09 2.09E-10 2.88E-09 7.47E-11 3502-T4 3.56E-04 7.02E-02 2.50E-05 5.81E-02 2.07E-05 7.87E-05 6.68E-08 1.97E-09 1.67E-12 1.63E-09 1.38E-12 -3.38E-10 -2.87E-13 Totals 3.31E-07 7.62E-09 3.34E-07 7.72E-09 3.16E-09 1.03E-10 Page 82 of 136

Enclosure 1 to ULNRC-05876 Table 4: Bin 12 IEF Scoping Analysis

[B.2-3 in 17671-014]

FAQ Affected Baseline 0048 Bin 12 Baseline Bin 12 Delta Fire IEF Bin 12 IEF CCDP CLERP LERF CDF Scenario (per yr.) Freq. (/yr.) CDF (per (per LERF CDF LERF

(/yr.) (per yr.) yr.) yr.) (per yr.) (per yr.) (per yr.)

4316-1 2.79E-05 1.32E-03 1.35E-03 6.14E-04 1.49E-05 1.71E-08 4.16E-10 8.28E-07 2.01E-08 8.11E-07 1.97E-08 4316-2 2.79E-05 1.32E-03 1.35E-03 6.14E-04 1.49E-05 1.71E-08 4.16E-10 8.28E-07 2.01E-08 8.11E-07 1.97E-08 4316-T1 4.67E-06 1.32E-03 1.32E-03 6.15E-04 1.49E-05 2.87E-09 6.97E-11 8.14E-07 1.98E-08 8.11E-07 1.97E-08 4316-T3 3.79E-04 1.32E-03 1.70E-03 4.49E-04 9.93E-06 1.70E-07 3.76E-09 7.63E-07 1.69E-08 5.93E-07 1.31E-08 3402-T1 5.67E-06 1.32E-03 1.33E-03 6.65E-04 1.73E-05 3.77E-09 9.79E-11 8.82E-07 2.29E-08 8.78E-07 2.28E-08 3402-T2 2.38E-04 1.32E-03 1.56E-03 6.15E-04 1.49E-05 1.46E-07 3.55E-09 9.58E-07 2.32E-08 8.11E-07 1.97E-08 3502-T2 7.78E-06 1.32E-03 1.33E-03 6.65E-04 1.73E-05 5.18E-09 1.34E-10 8.83E-07 2.29E-08 8.78E-07 2.28E-08 3502-T4 2.50E-05 1.32E-03 1.34E-03 7.87E-05 6.68E-08 1.97E-09 1.67E-12 1.06E-07 8.99E-11 1.04E-07 8.82E-11 Page 83 of 136 to ULNRC-05876 Probabilistic Risk Assessment RAI 08 Please clarify the following related to fire induced initiating events:

a. The NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities,"

apportionment method for weighting the influence factors for transient combustible ignition sources was designed to accommodate only integer values, although use of fractional values between the minimum of one and maximum of 10 (or 50 for "maintenance") is not precluded.

However, the only prescribed value below one is zero, as credit for administrative controls is considered to be already embedded in the transient fire frequencies based on the historical data.

Furthermore, a physical analysis unit with a total weight of zero would appear not to meet Supporting Requirement IGN-A9 in the ASME/ANS PRA Standard. The licensee's use of fractional values between zero and one would constitute a "deviation from 6850" for which at least a sensitivity analysis, using a minimum combined weight of one for the three influence factors, would be appropriate if any such locales have a combined weight less than one. Please provide a sensitivity study that shows the impact on the total and change in fire risk of using at least one weighting factor for Low (one) rather than the "special weighting factors."

b. The Ignition Frequencies Calculation, states that (1) the Callaway plant-specific fire history provided insufficient data for Bayesian update; (2) the generic fire frequencies are appropriate for Callaway; and (3) as a result, Bayesian update was not performed. Nonetheless, it appears that a reduced plant-specific value was used for Bin 16.2. Note that FAQ 35 (Supp. 1 of NUREG/CR-6850) states:

"In calculating the fire frequencies, the number of plant reactor years is based on the entire US fleet, i.e., it has been assumed that all existing plants contribute to the bus duct fire frequency."

This means that plants such as Callaway, with a lower number of iso-phase bus ducts than "typical," have already been, at least to some probably unquantifiable extent, implicitly included in the generic estimate. Therefore, the factor of five reduction is likely too generous.

Please provide a sensitivity analysis without this factor or an alternate approach to justify the use of such a factor.

Additional Justification Needed:

Include in the results the quantitative results, if any, on both delta-CDF and delta-LERF as well, since YD-1 has an associated VFDR. (Note: It appears that the answer is no change, since the VFDR does not appear to be related to the bus ducts, but this needs to be docketed to complete the response).

Response to Probabilistic Risk Assessment RAI-08

a. Response provided by ULNRC-05851 dated April 17, 2012.

b. Ignition bin 16.2 in NUREG/CR-6850, Supplement 1 involves iso-phase bus ducts. Callaway Plant uses considerably fewer iso-phase bus ducts than a typical plant. Callaway Plant preferentially uses Page 84 of 136 to ULNRC-05876 cable ducts for termination of high energy distribution points. In accordance with FAQ-0035, fires in cable ducts have been incorporated with the end device, thus they need no specific initiating event treatment. All of the Bin 16.2 components at Callaway Plant are contained within the fire area YD-1 Yard, specifically the Circulating Water Pump House (CWPH). Callaway Plant has 9 bus ducts, whereas the typical plant has approximately 45 bus duct components.

A sensitivity study was performed to quantify the risk impact of using the full bin 16.2 ignition frequency in the Callaway Plant FPRA. This sensitivity removes the factor of 5 reduction (i.e., 5 =

45/9) in the bin which was used in the Initiating Event Frequency Calculations (Calculation 17671-005) and recalculates the fire risk in the affected Fire Area(s). Table 1 shows the results of this sensitivity study. The line for LAR Values reflects the base case ignition source values in the CWPH area. The line for Full Bin 16.2 case reflects the full value for the bin 16.2 from NUREG/CR-6850 (Supplement 1). The results for each case, and the change in frequencies, are shown in the table below.

Table 1 - Bin 16.2 Full Value Sensitivity Results IEF CDF LERF CDF LERF Case (per yr.) CCDP CLERP (per yr.) (per yr.) (per yr.) (per yr.)

LAR 5.18E-03 1.24E-05 2.41E-07 6.42E-08 1.25E-09 0.00E+00 0.00E+00 Values Full Bin 5.84E-03 1.24E-05 2.41E-07 7.24E-08 1.41E-09 0.00E+00 0.00E+00 16.2 Frequency 6.60E-04 n/a n/a 8.18E-09 1.59E-10 0.00E+00 0.00E+00 Increase Fire area YD-1 Yard has a total Fire CDF of 1.03E-6/yr. and a total Fire LERF of 2.18E-8/yr. in the Callaway Plant NFPA 805 LAR submittal. Compared to the LAR risk totals, the CDF and LERF increases shown in Table 1 are less than 1% of each metric.

Fire Area YD-1 has a VFDR due to cables for the Refueling Water Storage Tank (RWST) water level sensors. These cables are not present in the Circulating Water Pumphouse, where all of the bus ducts are located. There are no VFDRs in the Circulating Water Pump House Area. There is no change in delta risk due to an increase in the IE frequency of bin 16.2. The increase in deterministic risk due to using the full bin 16.2 ignition frequency is considered to be negligible.

Page 85 of 136 to ULNRC-05876 Probabilistic Risk Assessment RAI 09 Please clarify the following issues related to uncertainty and sensitivity studies:

a. It was recently stated at the Nuclear Energy Institute Fire Protection Information Forum (NEI FPIF) that the Phenomena Identification and Ranking Table (PIRT) Panel being conducted for the DC circuit failure tests from the DESIREE-FIRE tests may be eliminating the credit (about a factor of two reduction) for control power transformers (CPTs) currently allowed by NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities,"

September 2005, as being invalid when estimating alternating current (AC) circuit failure probabilities. Please provide a sensitivity analysis that removes this CPT credit and the resulting impact on core damage frequency (CDF), LERF, delta-CDF (CDF), and LERF.

Please confirm that these potentially reduced probabilities based on CPT presence were not used to initially screen out components whose failure (or spurious operation) was due to fire-induced cable impacts from the subsequent analyses. Note also that assuming the presence of CPTs for control circuits in the MCR panels may be incorrect and, if so, should be removed when performing the sensitivity analysis.

Additional Justification Needed At the end of the response, the response to PRA RAI 13 is cited as a basis for concluding that the results in this RAI (9a) are bounding and conservative due to failure probabilities exceeding one. In the response to PRA RAI 13, was not the evaluation performed such that the effect of probabilities exceeding one was eliminated? Also, are all the effects from the RAI 13 Response included in the changes made in performing the sensitivity analysis for RAI 9a?

b. The Uncertainty and Sensitivity Analyses Calculation indicates that sensitivity/uncertainty analyses were not performed for fire ignition frequencies (other than the bins required by FAQ 48 in Supp. 1 to NUREG/CR-6850) or cable failure mode likelihoods. Please provide the results of sensitivity/uncertainty analyses for these values.

Additional Justification Needed When combining the risk and delta-risk increases per bin in Table 4, does the following correctly characterize how the whole-area burnup scenario contributions were included? For all the areas assumed to contribute per a single bin, e.g., A-28, et al., to bin 15.1, or grouped bin, e.g., A-3, et al., to grouped bin 5/6/7 in Table 3, the contribution from that area was included with the corresponding bin (or grouped bin) in Table 4. That is, while for scenarios with a single ignition source the contribution arose solely from the actual corresponding bin, for the whole-area burnup scenarios the contribution arose solely from the assumed corresponding bin, as per Table 3. If this is not the correct characterization, provide what is.

Page 86 of 136 to ULNRC-05876 Response to Probabilistic Risk Assessment RAI-09

a. The Callaway Plant FPRA did not use low spurious actuation probabilities to screen out components from inclusion in the FPRA. As such, the potentially increased spurious probabilities due to elimination of credit for control power transformers (CPT) have no effect on what components are included in the FPRA.

A sensitivity study has been performed to examine the effect of removing credit for CPTs in the calculation of spurious failure probabilities for AC powered control circuits. Merged, global risk equations were used, with appropriate manipulation of spurious failure probabilities, to estimate the impact on fire CDF, CDF, LERF, and LERF.

The impact on total CDF and total LERF is a straight forward process. All fire scenarios in the PRA are concatenated into a single equation and the failure probabilities are manipulated to increase the probabilities of certain basic events.

The impact on delta CDF/LERF is more difficult to calculate, so a surrogate measure is used.

Delta risk is the variant risk minus the compliant risk. Both variant risk and compliant risk can be represented by a cutset equation, which can be quantified to find the equation value. Delta risk is the difference between the variant risk value and compliant risk value. There is no cutset equation for delta risk. Delta risk is calculated during the Fire Risk Evaluation (FRE) process for each variant scenario. The total plant delta risk is the addition of delta risk from all variant scenarios. In order to calculate the delta risk exactly in this sensitivity study, it would be necessary to re-quantify each FRE.

As an alternative approach, all variant scenarios were concatenated into an equation and quantified. These merged, VFDR-only equations were modified with the appropriate increased spurious event probabilities and the increase in risk was assumed to be equal to the increase in delta risk. The change in risk for this equation was calculated and taken to be the maximum possible change in delta risk due to eliminating credit for CPTs.

The first step was to identify all AC powered components that credit a CPT for the calculation of spurious failure probabilities. Table 1 shows these components.

Table 1 Components with CPT Credit Equipment ID Component Type ALHV0005 MOV ALHV0007 MOV ALHV0009 MOV ALHV0011 MOV ALHV0030 MOV ALHV0031 MOV ALHV0032 MOV ALHV0033 MOV Page 87 of 136 to ULNRC-05876 Table 1 Components with CPT Credit Equipment ID Component Type ALHV0034 MOV ALHV0035 MOV ALHV0036 MOV BBHV0013 MOV BBHV0014 MOV BBHV0015 MOV BBHV0016 MOV BBHV8000A MOV BBHV8000B MOV BBHV8351A MOV BBHV8351B MOV BBHV8351C MOV BBHV8351D MOV BBPV8702A MOV BBPV8702B MOV BGHV8105 MOV BGHV8105 MOV BGHV8106 MOV BGHV8106 MOV BGHV8109 MOV BGHV8110 MOV BGHV8110 MOV BGHV8111 MOV BGHV8111 MOV BGHV8357A MOV BGHV8357B MOV BGLCV0112B MOV BGLCV0112C MOV BNHV8806A MOV BNHV8806B MOV BNHV8812A MOV BNHV8812B MOV BNHV8813 MOV BNLCV0112D MOV BNLCV0112E MOV EAHV0005 HDV EAHV0006 HDV EFHV0037 MOV Page 88 of 136 to ULNRC-05876 Table 1 Components with CPT Credit Equipment ID Component Type EFHV0038 MOV EFHV0051 MOV EFHV0052 MOV EFHV0059 MOV EFHV0060 MOV EFHV0066 MOV EGHV0011 MOV EGHV0012 MOV EGHV0013 MOV EGHV0014 MOV EGHV0015 MOV EGHV0015 MOV EGHV0016 MOV EGHV0016 MOV EGHV0053 MOV EGHV0053 MOV EGHV0054 MOV EGHV0054 MOV EGHV0058 MOV EGHV0061 MOV EGHV0062 MOV EGHV0069A AOV EGHV0069B AOV EGHV0070A AOV EGHV0070B AOV EGHV0071 MOV EJFCV0610 MOV EJFCV0611 MOV EJHV8701A MOV EJHV8701B MOV EJHV8716A MOV EJHV8716B MOV EJHV8804A MOV EJHV8804B MOV EJHV8809A MOV EJHV8809B MOV EJHV8811A MOV EJHV8811B MOV Page 89 of 136 to ULNRC-05876 Table 1 Components with CPT Credit Equipment ID Component Type EMHV8801A MOV EMHV8801B MOV EMHV8803A MOV EMHV8803B MOV EMHV8814A MOV EMHV8814B MOV EMHV8821A MOV EMHV8821B MOV EMHV8923A MOV EMHV8923B MOV ENHV0001 MOV ENHV0006 MOV ENHV0007 MOV ENHV0012 MOV EPHV8808A MOV EPHV8808B MOV EPHV8808C MOV EPHV8808D MOV LFFV0095 MOV NB0109 BKR NB0112 BKR NB0209 BKR NB0212 BKR PA0201 BKR VEA2101A HDV VEA2101B HDV Where: MOV = Motor Operated Valve AOV = Air Operated Valve BKR = Breaker HDV = Hydraulically Driven Valve The next step was to identify base case spurious failure probabilities associated with each component in Table 1. Since this sensitivity uses the global results equations for each risk metric, this list of basic events was trimmed down by only including spurious failure basic events that actually appear in at least one of the results equations. Table 2 shows:

a) all components appearing in the final risk equation (CDF or LERF) b) basic event name(s) for these components c) the base case value for spurious actuation of these components which credits the CPT Page 90 of 136 to ULNRC-05876 d) the sensitivity study value for spurious operation, which does not credit CPT.

Table 2 - Modified Spurious Failure Basic Events LAR PRA RAI 09a Equipment ID Spurious BE(s)

Value Probability ALHV0034 AL-MOV-SC-ALHV34 0.4 0.8 ALHV0035 AL-MOV-SC-ALHV35 0.4 0.8 ALHV0036 AL-MOV-SC-ALHV36 0.4 0.8 BB-MOV-SC-HV0013 0.4 0.8 BBHV0013 BB-MOV-SC-HV00133:00E-01 0.3 0.6 BB-MOV-SC-HV0014 0.4 0.8 BBHV0014 BB-MOV-SC-HV00143:00E-01 0.3 0.6 BB-MOV-SC-HV0015 0.4 0.8 BBHV0015 BB-MOV-SC-HV00153:00E-01 0.3 0.6 BB-MOV-SC-HV0016 0.4 0.8 BBHV0016 BB-MOV-SC-HV00163:00E-01 0.3 0.6 BBHV8000A BB-MOV-SC-8000A 0.4 0.8 BBHV8000B BB-MOV-SC-8000B 0.4 0.8 BB-MOV-SC-V8351A 0.4 0.4*

BBHV8351A BB-MOV-SC-V8351A3:00E-01 0.3 0.3*

BB-MOV-SC-V8351B 0.4 0.4*

BBHV8351B BB-MOV-SC-V8351B3:00E-01 0.3 0.3*

BB-MOV-SC-V8351C 0.4 0.4*

BBHV8351C BB-MOV-SC-V8351C3:00E-01 0.3 0.3*

BB-MOV-SC-V8351D 0.4 0.4*

BBHV8351D BB-MOV-SC-V8351D3:00E-01 0.3 0.3*

BG-MOV-SC-V112B 0.4 0.8 BGLCV0112B BG-MOV-SC-V112B1:00E-01 0.1 0.2 BGLCV0112C BG-MOV-SC-V112C 0.4 0.8 BNHV8806A BN-MOV-SC-V8806A 0.4 0.8 BNHV8806B BN-MOV-SC-V8806B 0.4 0.8 BNHV8813 BN-MOV-SC-HV8813 0.4 0.8 EA-HDV-SC-HV0005 0.4 0.8 EAHV0005 EA-HDV-SC-HV00053:00E-01 0.3 0.6 EA-HDV-SC-HV0006 0.4 0.8 EAHV0006 EA-HDV-SC-HV00063:00E-01 0.3 0.6 EFHV0051 EF-MOV-SC-EFHV51 0.4 0.8 EFHV0052 EF-MOV-SC-EFHV52 0.4 0.8 EF-MOV-SO-EFHV59 0.4 0.8 EFHV0059 EF-MOV-SO-EFHV593:00E-01 0.3 0.6 EFHV0060 EF-MOV-SO-EFHV60 0.4 0.8 Page 91 of 136 to ULNRC-05876 Table 2 - Modified Spurious Failure Basic Events LAR PRA RAI 09a Equipment ID Spurious BE(s)

Value Probability EF-MOV-SO-EFHV603:00E-01 0.3 0.6 EGHV0011 EG-MOV-SO-HV11 0.4 0.8 EGHV0012 EG-MOV-SO-HV12 0.4 0.8 EGHV0013 EG-MOV-SO-HV13 0.4 0.8 EGHV0014 EG-MOV-SO-HV14 0.4 0.8 EG-MOV-SC-EGHV58 0.4 0.8 EGHV0058 EG-MOV-SC-EGHV583:00E-01 0.3 0.6 EG-MOV-SC-EGHV61 0.4 0.8 EGHV0061 EG-MOV-SC-EGHV613:00E-01 0.3 0.6 EG-MOV-SC-EGHV62 0.4 0.8 EGHV0062 EG-MOV-SC-EGHV623:00E-01 0.3 0.6 EG-MOV-SC-HV00713:00E-01 0.3 0.6 EGHV0071 EG-MOV-SC-HV00714:00E-01 0.4 0.8 EJFCV0611 EJ-MOV-SC-FCV611 0.4 0.8 EJHV8716A EJ-MOV-SC-V8716A 0.4 0.8 EJHV8716B EJ-MOV-SC-V8716B 0.4 0.8 EJ-MOV-SO-V8811A 0.4 0.8 EJHV8811A EJ-MOV-SO-V8811A3:00E-01 0.3 0.6 EJ-MOV-SO-V8811B 0.4 0.8 EJHV8811B EJ-MOV-SO-V8811B3:00E-01 0.3 0.6 EM-MOV-SO-V8801A 0.4 0.8 EMHV8801A EM-MOV-SO-V8801A8:00E-01 0.8 0.8 EM-MOV-SO-V8801B 0.4 0.8 EMHV8801B EM-MOV-SO-V8801B8:00E-01 0.8 0.8 EMHV8803A EM-MOV-SO-V8803A 0.4 0.8 EMHV8803B EM-MOV-SO-V8803B 0.4 0.8 EMHV8814B- EM-MOV-SC-V8814B 0.4 0.8 PRA EM-MOV-SC-V8814B8:00E-01 0.8 0.8 EMHV8923A EM-MOV-SC-V8923A 0.4 0.8 ENHV0001 EN-MOV-SO-ENHV01 0.4 0.8 EN-MOV-SO-HV0006 0.4 0.8 ENHV0006 EN-MOV-SO-HV00063:00E-01 0.3 0.6 ENHV0007 EN-MOV-SO-ENHV07 0.4 0.8 EN-MOV-SO-HV0012 0.4 0.8 ENHV0012 EN-MOV-SO-HV00123:00E-01 0.3 0.6 EPHV8808A EP-MOV-SC-V8808A 0.4 0.8 EPHV8808B EP-MOV-SC-V8808B 0.4 0.8 EPHV8808C EP-MOV-SC-V8808C 0.4 0.8 Page 92 of 136 to ULNRC-05876 Table 2 - Modified Spurious Failure Basic Events LAR PRA RAI 09a Equipment ID Spurious BE(s)

Value Probability EPHV8808D EP-MOV-SC-V8808D 0.4 0.8 NB0212 NB-BKR-SC-NB0212 0.1 0.2 PA0201 PA-BKR-SO-PA0201 0.4 0.8

• No change. See explanation below for why these probabilities do not require change.

Explanation for BBHV8351A-D:

The probabilities for spurious closure of BBHV8351A, B, C and D were not increased in the sensitivity study, even though the base case values credit a CPT. The probabilities in the base case are high enough that they violate the rare event approximation for fault tree logic codes.

These four valves appear in an 'OR' gate for loss of seal injection flow. These four valves always appear in the same scenario. There are no scenarios with one or two valves. The current probability for loss of seal cooling when these basic events appear in a scenario is greater than

1. Increasing the probabilities for these valves would only increase the over-counting for loss of seal cooling.

There are two sets of probabilities that maximize seal cooling loss at a value of 1.0. These are:

a) each valve is assigned a .25, or b) one valve is assigned a 1.0 and the other 3 are assigned a 0.0.

For the sensitivity study, the valve probabilities for BBHV8351A-D were retained at the base case values of 0.3 and 0.4, which results in a total scenario probability for loss of seal cooling of 1.2 to 1.6.

As seen in Table 2, some events have multiple probabilities used for the same event. This occurs because in some instances, the component is only susceptible to internal hot shorts and in some cases the cable is also susceptible to external hot shorts. The 0.3 was used for fires in the main control board where external hot shorts were not considered valid. For the sensitivity study, all the base case (LAR) probabilities are doubled to create the value used in the sensitivity study.

The updated spurious failure probabilities were then imported into the global basic event data (BED) files for each of the four (CDF, CDF, LERF, and LERF) global risk equations. The resulting increase in each metric is shown in Table 3.

Page 93 of 136 to ULNRC-05876 Table 3 - Risk Metric Increases Risk Case Description Increase CDF (/yr) Estimate of plant-wide CDF increase 6.26E-06 LERF (/yr) Estimate of plant-wide LERF increase 1.48E-07 CDF-VFDR (/yr) Estimate of plant-wide CDF increase 1.75E-06 LERF-VFDR (/yr) Estimate of plant-wide LERF increase 7.36E-08 Table 4 provides some perspective by comparing these risk metric increases to their corresponding plant-wide totals reported in the license amendment request, and comparing the theoretical total of the two against the risk goals from Regulatory Guide (RG) 1.205.

Table 4 - Comparison of Risk Increases to Risk Goals LAR Plant Theoretical Metric Increase RG 1.205 Goal Total Total CDF (/yr) 6.26E-06 2.04E-05 2.67E-05 < 1E-4/yr.

LERF (/yr) 1.48E-07 3.97E-07 5.45E-07 < 1E-5/yr.

CDF (/yr) 1.75E-06 1.87E-06 3.62E-06 < 1E-5/yr.

LERF (/yr) 7.36E-08 3.84E-08 1.12E-07 < 1E-6/yr.

As shown, when the risk increases due to de-crediting the CPTs are added to the existing baseline risk metrics, the resultant theoretical totals are still below the risk goals presented in RG 1.205. In addition, increasing the probability of spurious operation to 0.8 for all MOVs and AOVs causes several PRA functions (such as loss of all RCP seal cooling, loss of CST inventory, loss of Steam Generator cooling) to become significantly greater than 1.0. This issue is discussed specifically for the BBHV8351A-D valves under Table 2 above and is also discussed in PRA RAI-13 to explain the generation of negative risk numbers. The issue involves using probabilities for basic events in an 'OR' gate that violate the rare event approximation, which is required by the WINNUPRA code for representative results. This issue has been identified and isolated to certain scenarios for the seal injection valves (BBHV8351A-D) and the CST drain valves (ADLV0079BA/BB), and is therefore possible to fix the issue with global PRA data changes. The issue is known to occur for other PRA functions, but it is not possible to isolate the issue to certain scenarios, so it is not possible to make global data changes to correct the issue. The over counting issue for the BBHV8351A-D valves was corrected for this sensitivity study. The over counting issue was not corrected for the CST drain scenario (valves ADLV0079BA/BB), nor any other PRA function. As such, the risk numbers presented in this sensitivity are considered conservative and bounding.

The issues from the PRA RAI-13 response (i.e., PORV and block valve operability) were not incorporated into this RAI sensitivity study. However, the effects of CPT credit (including retaining the BBHV8351A-D probabilities at the base case 0.4) were incorporated into PRA RAI-13.

Page 94 of 136 to ULNRC-05876

b. A sensitivity or uncertainty study on the cable failure likelihoods, beyond what was performed in part A of this RAI response, is not considered necessary. In part A of this RAI response, removing Control Power Transformer (CPT) credit in the cable failure likelihood calculations caused the vast majority of spurious failures in the Callaway Plant FPRA to have a value of 0.8. With so many values close to 1.0, there is little remaining uncertainty towards the upper bound of the uncertainty distribution, so the CPT de-crediting sensitivity in part A of this RAI response is considered sufficiently bounding. A sensitivity study was performed to consider the effects of ignition frequency uncertainty on fire risk, using the uncertainty parameters presented in NUREG/CR-6850, Supplement 1. The FPRA results, as presented in the Callaway Plant LAR, utilize the mean ignition frequency values presented in Supplement 1. In this sensitivity study, the ignition frequencies at the 95% confidence interval for each bin were used, and the effect on fire CDF, LERF, delta CDF (CDF), and delta LERF (LERF) was measured. The effects on these four risk metrics were calculated for each ignition frequency bin individually, and then all bins were combined to see the effect of using all bins at the 95%

confidence interval frequency simultaneously. This second study is more severe than a parametric uncertainty run using a Crystal Ball or a Monte Carlo approach, the final value of the sensitivity study assumes all fire frequencies are simultaneously at their 95% level.

Technical Approach The basis for the study was to multiply all scenario base line CDF/LERF values by a factor representing the multiplication factor for the 95th percentile bin ignition frequency. The first step was to determine the multiplication factor for each bin. This is simply the 95% frequency divided by the mean frequency. The multiplication factors for each bin are shown in Table 1.

Note that a special bin, 33/34/35, is defined for use with the catastrophic Turbine-Generator fire, which is dictated by the methodology in Appendix O of NUREG/CR-6850.

Table 1 - Ignition Frequency Bin Multipliers Mean value (used in 95th percentile Multiplier base case bin ignition used for Bin No. Location Ignition Source quant.) frequency study 1 Battery Room Batteries 3.26E-04 1.25E-03 3.83 Reactor Coolant 2 Containment (PWR) 2.35E-03 6.11E-03 2.60 Pump Transients and hot 3 Containment (PWR) 2.34E-03 5.89E-03 2.52 work Main Control 4 Control Room 8.24E-04 2.47E-03 3.00 Board Cable fires caused Control/Aux/Reactor 5 by welding and 1.25E-03 2.83E-03 2.26 Building cutting Page 95 of 136 to ULNRC-05876 Table 1 - Ignition Frequency Bin Multipliers Mean value (used in 95th percentile Multiplier base case bin ignition used for Bin No. Location Ignition Source quant.) frequency study Transient fires Control/Aux/Reactor 6 caused by welding 2.46E-03 5.65E-03 2.30 Building and cutting Control/Aux/Reactor 7 Transients 4.81E-03 8.80E-03 1.83 Building 8 Diesel Generator Room Diesel Generators 5.04E-03 9.02E-03 1.79 Plant-Wide 9 Air Compressors 4.65E-03 8.51E-03 1.83 Components Plant-Wide 10 Battery Chargers 1.18E-03 3.07E-03 2.60 Components Cable fires caused Plant-Wide 11 by welding and 9.43E-04 2.82E-03 2.99 Components cutting Plant-Wide Cable Run (Self-12 1.32E-03 3.43E-03 2.60 Components ignited cable fires)

Plant-Wide 13 Dryers 4.20E-04 1.61E-03 3.83 Components Plant-Wide 14 Electric motors 3.41E-03 6.61E-03 1.94 Components Plant-Wide Electrical Cabinets 15.1 2.36E-02 9.40E-02 3.98 Components Non-HEAF Plant-Wide Electrical 15.2 1.06E-03 2.75E-03 2.59 Components Cabinets-HEAF Plant-Wide 16.1 Bus Ducts 1.27E-03 3.31E-03 2.61 Components Plant-Wide Iso-phase Bus 16.2 1.65E-041 2.15E-03 13.03 Components Ducts Plant-Wide 17 Hydrogen Tanks 1.18E-03 3.07E-03 2.60 Components Plant-Wide 18 Junction box 1.11E-03 2.89E-03 2.60 Components Plant-Wide Misc. Hydrogen 19 1.24E-03 3.22E-03 2.60 Components Fires Off-gas/H2 Plant-Wide 20 Recombiner 8.83E-03 1.95E-02 2.21 Components (BWR)

Plant-Wide 21 Pumps 1.42E-02 2.06E-02 1.45 Components Page 96 of 136 to ULNRC-05876 Table 1 - Ignition Frequency Bin Multipliers Mean value (used in 95th percentile Multiplier base case bin ignition used for Bin No. Location Ignition Source quant.) frequency study Plant-Wide 22 RPS MG sets 9.33E-04 2.88E-03 3.09 Components Plant-Wide 23 Transformers 8.02E-03 1.29E-02 1.61 Components Transient fires Plant-Wide 24 caused by welding 3.65E-03 7.38E-03 2.02 Components and cutting Plant-Wide 25 Transients 8.28E-03 1.37E-02 1.65 Components Plant-Wide Ventilation 26 6.12E-03 1.04E-02 1.70 Components Subsystems Transformer -

27 Transformer Yard 1.62E-03 4.21E-03 2.60 Catastrophic Transformer - Non 28 Transformer Yard 8.38E-03 1.40E-02 1.67 Catastrophic Yard transformers 29 Transformer Yard 1.89E-03 3.79E-03 2.01 (Others) 30 Turbine Building Boiler 9.78E-04 2.55E-03 2.61 Cable fires caused 31 Turbine Building by welding and 4.50E-04 1.73E-03 3.84 cutting Main feedwater 32 Turbine Building 5.44E-03 1.00E-02 1.84 pumps Turbine Generator 33 Turbine Building 2.10E-03 4.98E-03 2.37 (T/G) Exciter 34 Turbine Building T/G Hydrogen 3.23E-03 6.79E-03 2.10 35 Turbine Building T/G Oil 3.89E-03 7.82E-03 2.01 Transient fires 36 Turbine Building caused by welding 7.55E-03 1.28E-02 1.70 and cutting 37 Turbine Building Transients 3.41E-03 7.07E-03 2.07 Catastrophic TG 33/34/35 Turbine Building 9.22E-03 1.96E-02 2.12 Fire Note 1 - The listed mean frequency for bin 16.2 is the NUREG/CR-6850 value reduced by a factor of 9/45 to account for the limited use of iso-phase bus ducts at Callaway Plant. This is the value used in the Callaway Plant LAR results. The 95th value is the full frequency from NUREG/CR-6850, Supplement 1.

Page 97 of 136 to ULNRC-05876 For transient fires at Callaway Plant, the room ignition frequencies are comprised of contributions from multiple bins. To simplify this sensitivity study and provide a conservative answer, the maximum multiplier from all transient bins for that area was used as a multiplier for the entire area fire transient initiating frequency. Bin 3 is an exception since transient fires in Containment have their own special bin. The multipliers used for transient fires are shown in Table 2.

Table 2 - Multipliers for Transient Ignition Frequencies Transient Bins Location Multiplier 5/6/7 Control/Aux/Reactor Building 2.30 31/36/37 Turbine Building 3.84 11/24/25 Plant-Wide Components 2.99 3 Containment (PWR) 2.52 Additionally, some fire areas were modeled as whole room burnup. The whole-room burnup scenarios use contributions from multiple ignition frequency bins to create a single, area-wide ignition frequency. To simplify this sensitivity study and provide a conservative answer, the highest single multiplier from all applicable bins for a given fire area were used as a multiplier for the entire area fire ignition frequency. A summary of the most limiting bin (MLB) in each whole-area burnout scenario is shown in Table 3.

Table 3 - Most Limiting Bin for Whole-Area Burnup Scenarios Fire Area Scenario MLB MLB Multiplier A-3 A3-WR 5/6/7 2.30 A-5 A5-WR 5/6/7 2.30 A-7 A7-WR 5/6/7 2.30 A-9 A9-WR 5/6/7 2.30 A-10 A10-WR 5/6/7 2.30 A-12 A12-WR 5/6/7 2.30 A-13 A13-WR 5/6/7 2.30 A-14 A14-WR 5/6/7 2.30 A-20 A20-WR 5/6/7 2.30 A-24 A24-WR 5/6/7 2.30 A-25 A25-WR 5/6/7 2.30 A-26 A26-WR 5/6/7 2.30 A-28 A28-WR 15.1 3.98 A-29 A29-WR 5/6/7 2.30 A-30 A30-WR 5/6/7 2.30 A-33 A33-WR 15.1 3.98 AB-1 AB1-WR 15.1 3.98 C-2 C2-WR 5/6/7 2.30 C-3 C3-WR 5/6/7 2.30 Page 98 of 136 to ULNRC-05876 Table 3 - Most Limiting Bin for Whole-Area Burnup Scenarios Fire Area Scenario MLB MLB Multiplier C-7 C7-WR 5/6/7 2.30 C-8 C8-WR 5/6/7 2.30 C-13 C13-WR 5/6/7 2.30 C-14 C14-WR 5/6/7 2.30 C-19 C19-WR 5/6/7 2.30 C-20 C20-WR 5/6/7 2.30 C-25 C25-WR 5/6/7 2.30 C-26 C26-WR 5/6/7 2.30 C-28 C28-WR 5/6/7 2.30 C-29 C29-WR 15.1 3.98 C-32 C32-WR 5/6/7 2.30 C-34 C34-WR 5/6/7 2.30 C-35 C35-WR 15.1 3.98 C-36 C36-WR 5/6/7 2.30 C-37 C37-WR 5/6/7 2.30 FB-1 FB-WR 15.1 3.98 LDF-1 LDF1-WR 15.1 3.98 RSB-1 RSB-WR 15.1 3.98 RW-1 RW-WR 15.1 3.98 UNCT UNCT-WR 15.1 3.98 UNPH UNPH-WR 15.1 3.98 USCT USCT-WR 15.1 3.98 USPH USPH-WR 15.1 3.98 YD-1 YD-SWYD 11/24/25 2.99 YD-1 YD-MXFR 11/24/25 2.99 YD-1 YD-SXFR 11/24/25 2.99 YD-1 YD-EX1 11/24/25 2.99 YD-1 YD-EX2 11/24/25 2.99 YD-1 YD-FPH 11/24/25 2.99 YD-1 YD-RWST 11/24/25 2.99 YD-1 YD-CST 11/24/25 2.99 YD-1 YD-CWPH 16.2 13.03 YD-1 YD-UHS 11/24/25 2.99 The CDF, CDF, LERF, and LERF results for each individual and whole-area burnup scenario were then increased with the appropriate multiplier (fixed scenario, transient scenario, or whole-area burnup scenario). The results are presented by bin, so that the increase in each metric per bin can be seen if the 95% confidence ignition frequency for that particular bin is used. Fire CDF and CDF results are shown in Table 4, and fire LERF and LERF are shown in Table 5. For the areas which only have a Page 99 of 136 to ULNRC-05876 single bin ignition frequency, or a grouped transient bin, from Table 3, the contribution from that area was included with the corresponding bin (or grouped bin) in Table 4. The transient bins and the catastrophic TG fire are not listed individually, but are listed as a group at the end of the table.

Table 4 - Fire CDF Results by Ignition Frequency Bin CDF LAR CDF Ignition RAI 9b RAI 9b Bin # LAR CDF Increase CDF Increase Source CDF CDF (/yr)

(/yr) (/yr) (/yr) 1 Batteries 6.09E-08 2.33E-07 1.72E-07 0.00E+00 0.00E+00 0.00E+00 Reactor 2 Coolant 4.19E-08 1.09E-07 6.70E-08 2.82E-10 7.33E-10 4.51E-10 Pump Transients 3 1.29E-07 3.24E-07 1.95E-07 1.54E-08 3.88E-08 2.34E-08 and hot work Main Control 4 6.64E-07 1.99E-06 1.33E-06 6.64E-07 1.99E-06 1.33E-06 Board Diesel 8 1.14E-08 2.03E-08 8.96E-09 0.00E+00 0.00E+00 0.00E+00 Generators Air 9 2.53E-09 4.63E-09 2.10E-09 0.00E+00 0.00E+00 0.00E+00 Compressors Battery 10 2.37E-07 6.16E-07 3.79E-07 0.00E+00 0.00E+00 0.00E+00 Chargers Cable Run 12 (Self-ignited 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 cable fires) 13 Dryers 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Electric 14 1.26E-08 2.45E-08 1.18E-08 1.64E-09 3.18E-09 1.54E-09 motors Electrical 15.1 Cabinets 9.78E-06 3.90E-05 2.92E-05 6.30E-07 2.51E-06 1.88E-06 Non-HEAF Electrical 15.2 Cabinets- 2.63E-07 6.83E-07 4.20E-07 3.66E-08 9.49E-08 5.83E-08 HEAF 16.1 Bus Ducts 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Iso-phase 16.2 6.42E-08 8.36E-07 7.72E-07 0.00E+00 0.00E+00 0.00E+00 Bus Ducts Hydrogen 17 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Tanks 18 Junction box 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Misc.

19 Hydrogen 2.79E-09 7.25E-09 4.46E-09 2.79E-09 7.25E-09 4.46E-09 Fires Page 100 of 136 to ULNRC-05876 Table 4 - Fire CDF Results by Ignition Frequency Bin CDF LAR CDF Ignition RAI 9b RAI 9b Bin # LAR CDF Increase CDF Increase Source CDF CDF (/yr)

(/yr) (/yr) (/yr)

Off-gas/ H2 20 Recombiner 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 (BWR) 21 Pumps 1.29E-06 1.87E-06 5.81E-07 1.66E-08 2.41E-08 7.47E-09 22 RPS MG sets 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 23 Transformers 2.64E-08 4.25E-08 1.61E-08 5.32E-09 8.56E-09 3.24E-09 Transformer 27 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

-Catastrophic Transformer 28 - Non 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Catastrophic Yard 29 transformers 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 (Others)

Ventilation 26 2.11E-07 3.58E-07 1.47E-07 5.25E-09 8.93E-09 3.67E-09 Subsystems 30 Boiler 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Main 32 feedwater 5.03E-07 9.24E-07 4.21E-07 0.00E+00 0.00E+00 0.00E+00 pumps Turbine 33 Generator 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 (T/G) Exciter T/G 34 8.46E-07 1.78E-06 9.33E-07 0.00E+00 0.00E+00 0.00E+00 Hydrogen 35 T/G Oil 3.06E-07 6.15E-07 3.09E-07 0.00E+00 0.00E+00 0.00E+00 33/34 Catastrophic 4.36E-07 9.27E-07 4.91E-07 0.00E+00 0.00E+00 0.00E+00

/35 TG Fire Combined CB/Aux/RB 5/6/7 3.11E-06 7.15E-06 4.03E-06 4.78E-07 1.10E-06 6.20E-07 Transient Bins Combined Turbine 31/36 Building 1.21E-06 4.64E-06 3.43E-06 0.00E+00 0.00E+00 0.00E+00

/37 Transient Bins Page 101 of 136 to ULNRC-05876 Table 4 - Fire CDF Results by Ignition Frequency Bin CDF LAR CDF Ignition RAI 9b RAI 9b Bin # LAR CDF Increase CDF Increase Source CDF CDF (/yr)

(/yr) (/yr) (/yr)

Combined 11/24 Plant-Wide 1.03E-06 3.09E-06 2.06E-06 1.68E-08 5.03E-08 3.35E-08

/25 Transient Bins Table 5 - Fire LERF Results by Ignition Frequency Bin RAI 9b LERF LAR RAI 9b LERF Ignition LAR Bin No. LERF Increase LERF LERF Increase Source LERF

(/yr) (/yr) (/yr) (/yr) (/yr) 1 Batteries 1.30E-09 4.98E-09 3.68E-09 0.00E+00 0.00E+00 0.00E+00 Reactor 2 Coolant 7.52E-10 1.95E-09 1.20E-09 7.58E-12 1.97E-11 1.21E-11 Pump Transients 3 8.07E-10 2.03E-09 1.22E-09 4.18E-10 1.05E-09 6.34E-10 and hot work Main 4 Control 1.76E-08 5.26E-08 3.51E-08 1.76E-08 5.26E-08 3.51E-08 Board Diesel 8 1.91E-12 3.43E-12 1.51E-12 0.00E+00 0.00E+00 0.00E+00 Generators Air 9 3.62E-12 6.63E-12 3.01E-12 0.00E+00 0.00E+00 0.00E+00 Compressors Battery 10 5.51E-09 1.43E-08 8.82E-09 0.00E+00 0.00E+00 0.00E+00 Chargers Cable Run 12 (Self-ignited 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 cable fires) 13 Dryers 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Electric 14 7.92E-11 1.54E-10 7.44E-11 4.45E-11 8.62E-11 4.17E-11 motors Electrical 15.1 Cabinets 1.62E-07 6.44E-07 4.83E-07 1.03E-08 4.11E-08 3.08E-08 Non-HEAF Electrical 15.2 Cabinets- 4.09E-09 1.06E-08 6.52E-09 5.23E-10 1.36E-09 8.35E-10 HEAF 16.1 Bus Ducts 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Iso-phase 16.2 1.25E-09 1.62E-08 1.50E-08 0.00E+00 0.00E+00 0.00E+00 Bus Ducts Page 102 of 136 to ULNRC-05876 Table 5 - Fire LERF Results by Ignition Frequency Bin RAI 9b LERF LAR RAI 9b LERF Ignition LAR Bin No. LERF Increase LERF LERF Increase Source LERF

(/yr) (/yr) (/yr) (/yr) (/yr)

Hydrogen 17 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Tanks 18 Junction box 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Misc.

19 Hydrogen 4.71E-13 1.22E-12 7.52E-13 4.71E-13 1.22E-12 7.52E-13 Fires Off-gas/H2 20 Recombiner 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 (BWR) 21 Pumps 2.34E-08 3.39E-08 1.05E-08 6.80E-11 9.87E-11 3.07E-11 RPS MG 22 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 sets Ventilation 26 3.65E-09 6.20E-09 2.55E-09 1.35E-10 2.29E-10 9.43E-11 Subsystems 23 Transformers 6.59E-10 1.06E-09 4.01E-10 3.47E-11 5.59E-11 2.11E-11 Transformer-27 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Catastrophic Transformer-28 Non 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Catastrophic Yard 29 transformers 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 (Others) 30 Boiler 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Main 32 feedwater 2.50E-09 4.59E-09 2.09E-09 0.00E+00 0.00E+00 0.00E+00 pumps Turbine Generator 33 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 (T/G)

Exciter T/G 34 2.20E-08 4.63E-08 2.43E-08 0.00E+00 0.00E+00 0.00E+00 Hydrogen 35 T/G Oil 7.81E-09 1.57E-08 7.89E-09 0.00E+00 0.00E+00 0.00E+00 Catastrophic 33/34/35 1.35E-08 2.86E-08 1.51E-08 0.00E+00 0.00E+00 0.00E+00 TG Fire Combined CB/Aux/RB 5/6/7 8.75E-08 2.01E-07 1.14E-07 9.29E-09 2.13E-08 1.21E-08 Transient Bins Page 103 of 136 to ULNRC-05876 Table 5 - Fire LERF Results by Ignition Frequency Bin RAI 9b LERF LAR RAI 9b LERF Ignition LAR Bin No. LERF Increase LERF LERF Increase Source LERF

(/yr) (/yr) (/yr) (/yr) (/yr)

Combined Turbine 31/36/37 Building 2.19E-08 8.40E-08 6.22E-08 0.00E+00 0.00E+00 0.00E+00 Transient Bins Combined Plant-Wide 11/24/25 2.18E-08 6.53E-08 4.35E-08 2.94E-12 8.81E-12 5.86E-12 Transient Bins Some bins may show zero risk contribution in Tables 4 and 5, even though sources in those bins exist at Callaway Plant and are included in the fire PRA. However, if a source only exists in areas that use whole-area burnup and its bin does not have the highest multiplier, then that contribution will not be shown here. For example, there are dryers (bin 13) at Callaway Plant in the Laundry Decontamination Facility (fire area LDF-1), but LDF-1 is a whole-area burnup and it contains some bin 15.1 components. Since bin 15.1 has the highest multiplier of any applicable bin in LDF-1, the bin 15.1 multiplier is used for the entire LDF-1 ignition frequency, and no contribution from the other, lower-multiplier bins, is shown. The bins that result in the highest increases in fire CDF, CDF, LERF, and LERF are summarized in Table 6. The theoretical total sums the increase in each metric with the associated plant-wide total in the current Callaway Plant LAR results.

Table 6 - Bins Causing the Highest Risk Metric Increases Largest LAR Plant- Theoretical Metric Increase Risk Goal Increase Wide Total Total Fire CDF (/yr) 15.1 2.92E-05 2.04E-05 4.96E-05 < 1E-4/yr.

Fire CDF (/yr) 15.1 1.88E-06 1.87E-06 3.75E-06 < 1E-5/yr.

Fire LERF (/yr) 15.1 4.83E-07 3.97E-07 8.80E-07 < 1E-5/yr.

Fire LERF (/yr) 4 3.51E-08 3.84E-08 7.35E-08 < 1E-6/yr.

Bin 15.1 has three of the four highest risk metric increases, which can be attributed to two primary causes. One, bin 15.1 has many risk-significant ignition sources, so increasing the frequency associated with those sources leads to a relatively large risk increase. Second, bin 15.1 has one of the highest multiplication factors of all ignition frequency bins. So, the combination of risk-significant sources and high multiplication factor lead to high contribution to risk increases in this sensitivity study.

Bin 4, which is the main control board, contributes notably to delta risk because the deterministically compliant risk in the main control room (fire area C-27) is assumed to be zero. As such, any risk increase also causes an equivalent increase in delta risk. Bin 4 also has a relatively high multiplication factor.

Page 104 of 136 to ULNRC-05876 As shown in Table 6, even if the bins that contribute the largest risk increases for each risk metric are added to the current plant-wide totals, the new, theoretical totals still have considerable margin to the risk goals.

The most extreme case in this sensitivity study is assuming that all ignition frequency bins use the 95% confidence frequency simultaneously. Table 7 shows this scenario, which is effectively a sum of all risk metric increases shown in Tables 4 and 5.

Table 7 - Risk Metric Increases with All Bins Using the 95% Frequency Total LAR Plant- Theoretical Metric Risk Goal Increase Wide Total Total Fire CDF (/yr) 4.48E-05 2.04E-05 6.52E-05 < 1E-4/yr.

Fire CDF (/yr) 3.96E-06 1.87E-06 5.83E-06 < 1E-5/yr.

Fire LERF (/yr) 8.34E-07 3.97E-07 1.23E-06 < 1E-5/yr.

Fire LERF (/yr) 7.96E-08 3.84E-08 1.18E-07 < 1E-6/yr.

As shown, the theoretical totals still show reasonable margin against the risk goals.

This sensitivity study increased each ignition frequency bin to its 95% confidence interval value and calculated the plant-wide risk metric increases. The increases were looked at individually (bin-by-bin),

as well as a bounding case, in which all bins were increased at the same time. Even when all ignition frequency bins are set to their 95% confidence interval ignition frequency, the plant risk metrics preserved adequate margin to the plant-wide risk goals. As such, it can be stated that the Callaway Plant fire PRA has reasonable assurance that statistical variations in the ignition frequency data will not cause significant changes in the risk insights or affect the decision-making processes associated with the transition to NFPA-805.

Page 105 of 136 to ULNRC-05876 Probabilistic Risk Assessment RAI 10 FAQ 52 (NUREG/CR-6850, Supp. 1) suggests growth times from zero to peak heat release rate (HRR) of 8 min and 2 min, respectively, for common trash type fires contained vs. uncontained within plastic or metal receptacles. These are based on Tests 7 through 9 of NUREG/CR-4860, "Flaw Density Examinations of a Clad Boiling Water Reactor Pressure Vessel Segment," February 1988 (the reference cited by Callaway in the MCR Fire Analysis Calculation as its basis for assuming a 10-min growth time [from which Callaway specifically cites Tests 3 and 4]), and the National Institute of Standards and Technology (NIST) and Lawrence Berkeley National Laboratory (LBL) tests. Please note that Tests 7 through 9 involved 5-gal and 30-gal polyethylene, unsealed trash cans containing clean cotton rags and paper, while Tests 3 and 4 involved a 2.5-gal polyethylene bucket containing "Kimwipes" and acetone. Thus, it would appear Tests 7 through 9 were more representative of the type of trash can fire to be expected in a minimal maintenance locale such as the MCR, while Tests 3 and 4, cited by the licensee as the basis for the longer growth time to maximum HRR, were more representative of the type of trash can fire to be expected in at least an occasional maintenance locale.

For Tests 7 through 9, the FAQ cites times to initial peak in fire intensity of 7, 8, and 13 min, respectively (i.e., two of the three cited tests support the recommended time of 8 min). Please provide the basis for the assumption of the applicability of Tests 3 and 4, such that the longer 10-min growth time was assumed, including a quantitative estimate of the effect of assuming the appropriate shorter growth time(s).

Additional Justification Needed The repeated claim that use of 8 min as a best estimate for the time to reach maximum HRR for a trash can fire is overly conservative should be removed from the response, as it is contradicted by the evidence provided in the RAI, based on FAQ 08-0052. Note that the shift of the time from 10 to 8 min (only a 20% effect) caused the probabilities of abandoning the MCR to increase from 50% to 135%. Focus the response on the sensitivity evaluation performed, which indicates a very small increase in CDF and should bound the increase in delta-CDF.

Response to Probabilistic Risk Assessment RAI-10

[Note the reference above to NUREG/CR-4860 is meant to refer to NUREG/CR-4680, Heat and Mass Release for Some Transient Fuel Source Fires: A Test Report (NUREG/CR-4680, SAND86-0312)]

Transient fires in the main control room are more likely to be associated with contained trash can fires than plastic bags (where the latter would be more applicable to Auxiliary Building areas). In this case the applicable FAQ recommended time to maximum HRR is 8 minutes. However, using the data in NUREG/CR-6850 Supplement 1 for the tests that are cited, a best estimate time of 10 minutes was derived and used in the LAR analysis.

In response to this RAI, sensitivity runs were performed on the main control room transient fires to investigate the impact on abandonment time for transient fire reaching peak HRR in 8 min as recommended in FAQ-0052.

Page 106 of 136 to ULNRC-05876 A transient fire was postulated for the electrical cabinet area (ECA) and the main control area (MCA).

CFAST runs were performed to find the probability of forced evacuation prior to suppression for the case with ventilation operable and with ventilation failed. The probability of forced evacuation using an 8 minute and 10 minute time for achieving peak heat release rate are shown in Table 1. Table 2 shows the change in CDF when the 8 minute time is used instead of the 10 minute time.

Table 1 - Difference in Evacuation Probabilities for 8 and 10 Minute Peak HRR Probability of Probability of Forced Forced Evacuation using Evacuation using 10 min to peak 8 min to peak Fire Case (Transient only) HRR HRR Transient fire in MCA - ventilation failed 7.79E-4 1.16E-3 Transient fire in MCA - ventilation operable 3.95E-5 9.39E-5 Transient fire in ECA - ventilation failed 0.0 0.0 Transient fire in ECA - ventilation operable 0.0 0.0 This sensitivity study applies only to transient fires. Transient fires originating in the ECA are shown to be of insufficient strength to cause evacuation. The combination of a high ceiling and limited combustible content lead to a situation where it is not possible to create a hot gas layer or cause opacity restrictions. Thus, the sensitivity study has no effect on transient fires occurring in the ECA.

The change in CDF using the 8 min peak HRR as opposed to the 10 min HRR is shown in Table 2.

Table 2 - Increase in CDF Caused by Using 8 min Peak HRR Versus 10 min Peak HRR Case Change in CDF Comment

(/yr)

MCA-vent operable +7.5E-10 Change is a factor of 135%, but absolute value of CDF is very small.

MCA-vent failed 0.0 Transient fire in the MCA will not damage any control panel that affects control room ventilation.

ECA-vent operable 0.0 Due to the volume of the ECA, a transient fire of typical combustible volume will not cause uninhabitable conditions prior to suppression.

ECA-vent failed 0.0 Due to the volume of the ECA, a transient fire of typical combustible volume will not cause uninhabitable conditions prior to suppression.

Since transient fires in the ECA cannot lead to evacuation, regardless of the HRR, the change in assumption of peak HRR does not change CDF results. For transient fires in the MCA, only the case Page 107 of 136 to ULNRC-05876 with ventilation operable is used. The control cabinets for the CR ventilation are in the ECA and it was assumed the CR ventilation remains operable for transient fires in the MCA. This is a conservative assumption, because the increase in evacuation probability is greatest for the ventilation operable case.

The change in CDF when using an 8 minutes peak HRR rather than a 10 min peak HRR is +7.5E-10/yr. This increase is insignificant when compared to the entire control room CDF of 7.8E-7/yr.

Probabilistic Risk Assessment RAI 11 Attachment G of the LAR identifies the ASP (RP118B) as a Primary Control Station (PCS). There is then a continuation of a bulleted list which includes numerous indications and controls which are also identified as PCSs. Please clarify that there are no other ex-control room locations (other than RP118B) considered as a PCS and that all the instruments and controls in the list are on RP118B.

Otherwise, please explain the apparent discrepancy.

Response to Probabilistic Risk Assessment RAI-11 Response provided by ULNRC-05851 dated April 17, 2012.

Page 108 of 136 to ULNRC-05876 Probabilistic Risk Assessment RAI 12 Area C-10 includes recovery actions to isolate Reactor Coolant System (RCS) injection flow to avoid pressurizer Power-Operated Relief Valve (PORV) challenge on pressurizer overfill. The spurious injection flow path involves high pressure safety injection flow path. During the audit, the licensee identified that plant-specific calculations determined that about 36 minutes are available to isolate the flow path prior to reaching water solid conditions in the pressurizer. This time seems to be longer than reasonably expected. (The NRC staff notes that FSAR Section 15.5.1.2 states that the pressurizer is water solid following a spurious SI signal at 8.75 minutes, even assuming the operator terminates normal charging pump flow at 6 minutes.)

Please provide the details of the calculation to justify that 36 minutes is available prior to water solid conditions, including assumptions related to assumed automatic pressure control response of the pressurizer spray valves and relief valves, the status of RCS letdown paths, and assumed operator responses, to justify the difference between the safety analysis of spurious SI and this scenario. In addition, please provide the details of the calculation of the human error probability which describe the basis for the time available to perform the action compared to the time to access the manual valve and close it to confirm this action is feasible. The response should justify the assumptions made to bound the time, and the assumptions as to the procedural response to a spuriously open injection flow path (i.e., is a manual actuation of Emergency Core Cooling System (ECCS) required which may further delay the recovery action).

Additional Justification Needed Can the MAAP run be conducted using the initial conditions from the FSAR 15.5.1.2 analysis, or the RETRAN run using the initial conditions from the MAAP analysis for S-20? If so, compare the results to show that the MAAP run is the more accurate representation of the actual scenario. The fact that the FSAR calculation may be more conservative does not necessarily account for the significant difference in timing (~27 min).

Response to Probabilistic Risk Assessment RAI-12 Description of PRA Scenario The fire scenario in question involves a fire in the B train class IE switchgear room (fire area C-10) which causes a spurious safety injection signal while at the same time failing the control circuits for the B train Centrifugal Charging Pump (CCP). Both Pressurizer Power Operated Relief Valves (PORVs) are simultaneously failed closed by fire damage. The CCPs will continue to inject, causing pressurizer pressure and level to increase. The pressurizer safety valves may open during the transient to relieve pressure, but initially they will relieve steam and reclose. Eventually, the pressurizer will become water solid and the safety valves will pass water. The safety valve is assumed to fail open once it passes water. This condition becomes an S2 LOCA which requires ECCS injection and recirculation in order to prevent core uncovery. The timing for this sequence was derived from a MAAP run (S-20 in the project MAAP report FAI/10-504 Callaway Plant MAAP 4.0.7 Fire PRA Sequences).

Page 109 of 136 to ULNRC-05876 Description of MAAP Analysis The salient features of the MAAP run S-20, from which the timing of this FPRA scenario was derived, are shown in Table 1 below:

Table 1 - Parameters for MAAP Run and FPRA Scenario Parameter MAAP Run S-20 C-10 FPRA Scenario Initiator Spurious Actuation of Spurious SI Signal at T=0 ECCS at T=0 Reactor trip T=0 T=0 Number of Injecting 2, through boron injection 2, through boron injection Charging pumps header header PORV status Operable Failed Letdown Isolated at 5 min Isolated at T=0 Normal charging pump Not modeled Tripped at T=0 Pressurizer heaters Operating Off Pressurizer sprays Not credited Not credited Time to PORV Opening 4 min None The S-20 MAAP run is designed to show the time the pressurizer would go water solid for scenarios with unmitigated safety injection in the absence of a LOCA event. As seen in Table 1, the MAAP S-20 scenario does not replicate all aspects of the actual FPRA scenario in C-10. In order to be efficient in generating MAAP analysis, the MAAP run S-20 applies to multiple fire scenarios. The MAAP S-20 scenario however, is conservative with respect to the C-10 FPRA scenario, principally because the PORVs are functional in the MAAP run as opposed to failed in the C-10 PRA scenario. This causes a higher reactor pressure throughout the scenario which leads to a lower ECCS flowrate. The MAAP run S-20 assumed the PORVs were operable, so that steam relief from the pressurizer occurred at the set point of the PORVs (2335 psig) rather than the set point of the safety valves (2425 psig). The pressurizer pressure would be higher during the actual FPRA scenario (with the failed PORVs),

which would result in a lower charging pump flow rate and a longer time to go water solid, than in MAAP Run S-20.

Explanation of MAAP Results Table 2 explains the critical MAAP features in order to provide an understanding of the results.

Page 110 of 136 to ULNRC-05876 Table 2 - MAAP Results for S-20 Parameter Value Total Pressurizer Volume 1854 ft3 Pressurizer Vapor Space during Normal Power Operation 804 ft3 Average ECCS Flowrate (from S-20 results Fig 3-701) 135 gpm Average ECCS Flowrate (from S-20 results Fig 3-702) 126 gpm Normal Letdown Flow 120 gpm Specific volume of water at 652F (pzr water temp) .02691 Total RWST water outflow, from Fig 3-702 4615 gallons Time to fill pressurizer 36.6 minutes Fig 3-701 shows ESF flowrate in lbm/hr, on a scale of 0-83,000. Fig 3-702 shows RWST water level in feet, on a scale of 38.9 feet to 40.3 feet. These two figures are used to calculate how much water is injected into the RCS during the transient. The charging pumps are secured at 36.6 minutes in the MAAP run. The flow rates in Table 2 above are interpolations from the figures. Due to uncertainty in reading numbers off a graph, the two figures produce differing values for ECCS flow rate. The outflow of the RWST should be the same as the ECCS flow rate. The volumes and flow rates were converted to gpm assuming properties of water at 100F. The RWST water level curve was deemed more accurate, so the calculation assumes ECCS flow rate was 126 gpm.

The free volume of the pressurizer is 804 ft3. The total net inflow into the RCS is 4015 gallons. (This is 4615 gal - minus the 600 gallons that escaped the letdown line until isolated at 5 minutes). The increase in water level occurs in the pressurizer where the temperature is 652F and the specific volume of water is .02691. The additional volume of the 4015 gallons is:

(4015 gal) * (8.3 lb./gal) * (.02691 specific vol.) = 897 ft3.

The available free volume of the pressurizer is 804 ft3, which results in a surplus water volume of 93 ft3. This is interpreted that 93 ft3 exited the pressurizer PORV when cycling prior to the pressurizer filling with water.

There are configuration differences between the MAAP run S-20 and the fire PRA scenario. These are shown in Table 3, with a qualitative assessment of the effect on the results.

Page 111 of 136 to ULNRC-05876 Table 3 - Scenario Differences between MAAP Run S-20 and FPRA MAAP FPRA Parameter Run scenario Effect on Result Pressurizer On Off The pressurizer heaters will cause the pressure to be higher, heaters thereby reducing charging flow. The MAAP run will produce an optimistic result for the FPRA scenario.

PORV Operable Failed The pressurizer is controlled at 2335 psig in MAAP, rather than 2425 psig in the PRA scenario. The MAAP flow rate will be higher, thus providing a shorter time to pressurizer fill. The MAAP run will produce a conservative result for the FPRA.

Letdown Isolated Isolate at Letdown flow is 120 gpm at 120F. By allowing letdown flow to at T=5 T=0 continue for 5 minutes in MAAP, the RCS inventory balance is reduced by 600 gallons. The free surface of the RCS is in the pressurizer which operates at a temperature of 652F, the temperature condition at which the 600 gallons must be reconciled. By not isolating letdown until 5 minutes, the MAAP run artificially provides an additional 133 ft3 of pressurizer space. Adjusting the calculation to account for this space would shorten the time to water solid by 133/28.1 = 4.7 minutes. (28.1 ft3 is the volumetric flow rate of 120 gpm at pressurizer water conditions).

Comparison with FSAR Analysis FSAR Chapter 15.5.1.2 describes the plant response to an inadvertent ECCS injection, caused by a spurious SI signal, which shows the pressurizer goes solid in 8.6 minutes. The RAI questions why the MAAP result is different from the FSAR case in Chapter 15.5.1.2.

The FSAR analysis is a licensing calculation which uses the RETRAN code and uses conservative initial conditions. The MAAP code is a best estimate code and uses best estimate initial conditions. A comparison of salient parameters and results between the FSAR run and the MAAP run are shown in Table 4.

Page 112 of 136 to ULNRC-05876 Table 4 - Comparison of MAAP and RETRAN Parameters Parameter MAAP Run S-20 FSAR 15.5.1.2 Initiator Spurious Actuation of Spurious SI Signal at T=0 ECCS at T=0 Reactor trip T=0 T=0 Number of injecting 2, through boron 2, through boron injection header. Normal Charging pumps injection header charging pump for first 6 minutes PORV status Operational Operational after 9 minutes Letdown Isolated at 5 min N/A Normal charging pump Not in model Operating for 6 minutes Pressurizer heaters Operating Off Pressurizer sprays Not credited Operating Pressurizer water solid 36.6 min 8.75 min Nominal flow rate into 126 gpm 346 gpm for 6 minutes RCS 299 gpm thereafter The differences between FSAR modeling and MAAP modeling cannot be compared on a specific basis without recourse to each code with the ability to do sensitivity studies. The decisive difference is the ECCS flow rate. These dictate pressurizer fill rate. The MAAP run is 126 gpm versus the 346 gpm for RETRAN. It is emphasized that the MAAP run is a best estimate calculation while the FSAR calculation is intentionally conservative.

Operator Action The NFPA 805 recovery action human error probability (HEP) in this sequence is calculated to be 1.9E-2. This action was walked down and shown to be feasible and accessible within the time constraints of the scenario. The HRA shows a time delay of 25 minutes until the procedural cue is reached. Execution time is 7 minutes, which includes 2 minutes of travel time. No recovery is credited on either the cognitive or execution portion of the HEP.

Sensitivity Study with No Recovery Action It is not feasible to run comparison studies of this scenario in MAAP and RETRAN. The codes are sufficiently different that the final numbers would not be expected to be similar, even for identical initial reactor conditions. Therefore, as a final aspect of the analysis, a sensitivity study was performed which removed credit for this recovery action in all fire scenarios where it is credited. The results are shown in Table 5. The changes in CDF/LERF/delta CDF/delta LERF are small and do not affect the overall risk profile.

Table 5 - Change in CDF/LERF with No Recovery Action Metric CDF (/yr) LERF (/yr)

Change in CDF/LERF +1.40E-7 +2.10E-9 Change in CDF/LERF +1.28E-7 +1.90E-9 Page 113 of 136 to ULNRC-05876 Probabilistic Risk Assessment RAI 13 During the audit, the licensee stated that there are some fire scenarios (e.g., in Areas C-21, C-22, and C-24) where a single fire could cause spurious opening of a PORV as well as the loss of power required to close the associated block valve. Isolation of this leak path requires that operators cause the PORV to close by locally opening its direct current (DC) breaker. Please discuss the fire scenarios which cause this failure mode which would require a local operator action to restore RCS integrity.

The response should address the frequency of fire scenarios, a description of the scenario, the locations of the target cables in terms of physical separation between the fire source and the two targets, and the total risk reduction which would be available if this failure mode were eliminated. In addition, please describe the operator recovery action in terms of its complexity, the time available to complete the action before reaching an unrecoverable condition, and in the context of each fire scenario with regards to other local recovery actions which might be required. A discussion of the risk importance of this recovery action should also be provided in terms of the change in risk if the action were assumed to be unsuccessful.

Additional Justification Needed Do the calculational results (Tables 2 through 7) reflect removal of credit for CPTs in all scenarios?

The discussion following Table 5 suggests that some CPT credit was retained (i.e., use of 0.4 for spurious operation probability). How would the results change if this credit were removed entirely?

Note that, for the delta calculations, the credit should be removed in both the base and comparative cases.

Response to Probabilistic Risk Assessment RAI-13 The RAI has four specific parts as indicated below. Each aspect is discussed in its own section. Parts 2 and 4 require a sensitivity study. The sensitivity study was performed with and without credit for CPTs in probability of spurious operation (issue from PRA RAI-9a). Probabilities for spurious closure of BBHV8351A-D were retained at their base case values in this sensitivity study as in PRA RAI-9a. Responses to part 2 and 4 therefore have two sets of sensitivity results.

1) The response should address the frequency of fire scenarios, a description of the scenario, the locations of the target cables in terms of physical separation between the fire source and the two targets.

2) Total risk reduction which would be available if this failure mode was eliminated.

3) Please describe the operator recovery action in terms of its complexity, the time available to complete the action before reaching an unrecoverable condition, and in the context of each fire scenario with regards to other local recovery actions which might be required.

4) A discussion of the risk importance of this recovery action should also be provided in terms of the change in risk if the action were assumed to be unsuccessful.

Page 114 of 136 to ULNRC-05876 Part 1: Description of Scenario and Where it Occurs The pressurizer has two power operated relief valves (PORV). The PORV is a DC power operated valve. Each PORV discharges to the Pressurizer Relief Tank through a dedicated line. Each PORV line is supplied with a motor operated isolation valve (i.e. block valve), which is powered by 480vac power. The power for the PORV and the block valve are from the same train. Thus, cables for the PORV and the block valve on the same discharge line are routed in the same fire areas. If sufficient cabling exists in a fire area such that a Pressurizer PORV could fail spuriously open and its associated block valve could also fail as-is open, a VFDR was assigned to that fire area in the Nuclear Safety Capability Assessment (NSCA) analysis. Fire Areas with such a VFDR are listed in Table 1.

Table 1 - Fire Areas with PORV-LOCA VFDRs Fire Area VFDR-ID Recovery Action Credited A-8 A-08-003 No A-11 A-11-001 No A-16 A-16-SOUTH-001 No A-17 A-17-001 No A-18 A-18-003 No A-27 A-27-003 No C-18 C-18-002 No C-21 C-21-003 YES C-22 C-22-003 YES C-23 C-23-002 No C-24 C-24-002 YES C-27 C-27-026 YES C-27 C-27-029 YES C-30 C-30-002 No C-33 C-33-002 No RB-1 RB-03-001 / RB-05-001 YES The PORVs are designed to fail closed on loss of DC power. The recovery action (RA) assigned to these VFDRs is to locally de-energize the PORV control circuit. A recovery action is not being credited in all fire areas in which the VFDR occurs as further explained below. For the non-RA areas, the fire risk quantification reflects the risk of not recovering RCS integrity for fires that cause spurious PORV actuation and block valve failure. The risk has been shown to be acceptably low, which can be attributed to one of several reasons:

Page 115 of 136 to ULNRC-05876 a) VFDRs are assigned on a Fire Area wide basis. The use of fire modeling would likely demonstrate that few, or even no, fire scenarios have sufficient cable damage to lead to the VFDR.

b) For fire scenarios which have the VFDR, sufficient plant systems to mitigate the PORV-LOCA are available and the risk of core damage is very low.

c) Due to low fire frequency and fixed suppression systems, the initiating event frequency for the scenario is very low.

The fire initiating event frequency for those scenarios which have the PORV-VFDR are shown in Table 2.

Table 2 - Initiating Event Frequency of Fire Scenarios with PORV-VFDR Scenario Initiator Fire Frequency Area Fire Scenario (per year) Scenario Description1 A-8 (none)2 n/a n/a A-11 1335-3 9.94E-08 Transient Fire (suppression failed) 2 A-16 (none) n/a n/a A-17 A17TS1 1.33E-05 Transient Fire A-17 A17TS4 1.27E-06 Transient Fire A-18 1410-3 9.01E-07 RJ159 (BOP computer)

A-27 (none) 2 n/a n/a C-18 3419-6 5.95E-07 Transient Fire C-18 3419-8 5.95E-07 Transient Fire C-18 3419-9 1.09E-07 Transient Fire (suppression failed)

C-21 3501T15 5.03E-08 Transient Fire C-21 3501T16 7.76E-09 Transient Fire C-21 3501T18 1.92E-07 Transient Fire C-21 3501TXX 8.46E-08 Transient Fire (suppression failed)

C-22 3801T2 4.17E-07 Transient Fire C-22 3801T3 5.83E-07 Transient Fire C-22 3801T10 6.38E-08 Transient Fire C-22 3801T14 1.25E-07 Transient Fire C-22 3801TXX 1.25E-07 Transient Fire (suppression failed)

C-23 3505-5 1.09E-07 Transient Fire (suppression failed)

C-24 3504-1 1.19E-06 Transient Fire C-24 3504-3 1.19E-06 Transient Fire C-24 3504-5 1.09E-07 Transient Fire (suppression failed)

Page 116 of 136 to ULNRC-05876 Table 2 - Initiating Event Frequency of Fire Scenarios with PORV-VFDR Scenario Initiator Fire Frequency Area Fire Scenario (per year) Scenario Description1 C-27 RL001/02 3.46E-06 MCB RL1/2 (No Evacuation)

C-27 RL021/022 3.46E-06 MCB RL21/22 (No Evacuation)

MCB Panel Propagation Scenario C-27 RL1-2-3-4 3.98E-07 (No Evacuation)

MCB Panel Propagation Scenario C-27 RL21-22-23-24 3.98E-07 (No Evacuation)

C-30 3617-4 1.19E-06 Transient Fire C-30 3617-5 1.09E-07 Transient Fire (suppression failed)

C-33 3804-6 3.44E-07 Transient Fire C-33 3804-7 3.44E-07 Transient Fire C-33 3804-19 1.09E-07 Transient Fire (suppression failed)

RB-1 RB3-T1 2.89E-05 Transient Fire RB-1 RB3-T2 1.88E-05 Transient Fire RB-1 RB5-T6 3.23E-06 Transient Fire RB-1 RB5-T7 4.26E-06 Transient Fire Totals 8.62E-05/yr.

Note 1 - A detailed description of each fire is provided in Attachment 6 of the Detailed Fire Modeling Report for each Fire Area Note 2 - No fire scenarios were identified that have the VFDR As noted, a detailed description of each fire scenario is provided in Attachment 6 of the detailed fire modeling report for each fire area. Attachment 8 of the same document lists the targets for each fire scenario and the time to damage for each target. The targets listed in Attachment 8 are raceways, and the cables in each raceway can be found in the Nuclear Safety Capability Analysis (Calculation KC-26), as well as the SAFE database.

Part 2: Risk Reduction Worth if the PORV-Block Valve Commonality Were Eliminated:

Table 3 shows the CDF and delta CDF for each scenario. The first CDF column (called Scenario CDF with VFDR [RA credited if applicable]) shows the CDF of the post-transition plant. This includes core damage arising from the VFDR, with the recovery action (if applicable as shown in Table 1) plus the core damage frequency contribution from all other failures in the scenario. The second CDF column (called Scenario CDF of complaint case [no VFDR]) shows the CDF for the scenario if all VFDRs are eliminated, (i.e. the PORV-VFDR and all others). The third CDF column shows the difference between the two, which is delta CDF.

Page 117 of 136 to ULNRC-05876 For this sensitivity study, the compliant case CDF is used as a bounding estimate of the risk reduction if only the PORV LOCA VFDR failure mode was removed. Since the compliant case removes all VFDRs, it reduces risk to a greater extent than if only the PORV LOCA VFDRs were removed.

If a recovery action is credited for any fire scenario in a fire area, it is credited for all fire scenarios in the fire area.

Note that all of the results tables have an a and a b version due to the combination of this issue with the issue from RAI 9a. The a versions show risk results that credit control power transformers (CPT) during the calculation of spurious actuation probabilities based on fire-induced cable damage.

The b version of each table show risk results with the CPT credit removed. Tables 3a and 3b show the fire CDF reduction if the PORV VFDR failure mode was removed from all scenarios in which it exists.

Table 3a - Bounding CDF Risk Reduction if PORV Failure Mode was Eliminated (with CPT)

CDF with VFDR CDF of

[RA credited if Complaint Case applicable] [no VFDR] CDF Reduction Fire Scenarios (per year) (per year) (per year)

All with PORV VFDR 3.37E-06 3.23E-06 1.31E-07 Table 3b - Bounding CDF Risk Reduction if PORV Failure Mode was Eliminated (no CPT)

CDF with VFDR CDF of

[RA credited if Complaint Case applicable] [no VFDR] CDF Reduction Fire Scenarios (per year) (per year) (per year)

All with PORV VFDR 4.77E-06 4.64E-06 1.36E-07 As shown, the total reduction in CDF if the PORV LOCA VFDR were eliminated from the fire areas is approximately 1.31E-7/yr. with CPT credit and 1.36E-7/yr. without CPT credit. Compared to the current plant-wide fire CDF of approximately 2.04E-5/yr. these reductions are not significant. The current plant-wide fire delta CDF is 1.87E-6/yr. ,which has considerable margin to the Regulatory Guide 1.205 limit of 1E-5/yr. The difference between the a and b results for delta risk is very small, because the CPT credit was removed from both the variant condition and the compliant condition.

Tables 4a and 4b shows the LERF equivalent of the results in Tables 3a and 3b.

Page 118 of 136 to ULNRC-05876 Table 4a - Bounding LERF Risk Reduction if PORV Failure Mode was Eliminated (with CPT)

LERF with VFDR LERF of Complaint LERF

[RA credited if Case [no VFDR] Reduction Fire Scenarios applicable] (per year) (per year) (per year)

All with PORV VFDR 1.91E-07 1.88E-07 3.60E-09 Table 4b - Bounding LERF Risk Reduction if PORV Failure Mode was Eliminated (no CPT)

LERF with VFDR LERF of Complaint LERF

[RA credited if Case [no VFDR] Reduction Fire Scenarios applicable] (per year) (per year) (per year)

All with PORV VFDR 2.47E-07 2.43E-07 3.80E-09 As shown in Table 4, the total reduction in LERF if the PORV LOCA VFDR were eliminated from the fire areas is approximately 3.60E-9/yr. with CPT credit and 3.80E-9/yr. without CPT credit. Compared to the current, plant-wide fire LERF of approximately 3.97E-7/yr., these reductions are not significant.

The current plant-wide fire delta LERF is 3.84E-8/yr., which has considerable margin to the Regulatory Guide 1.205 limit of 1E-6/yr. The difference between the a and b results for delta risk is very small, because the CPT credit was removed from both the variant condition and the compliant condition.

Part 3: Description of Operator Recovery Action:

The operator recovery action for the selected VFDRs involves removal of DC power from the PORV control circuit for BBPCV0455A or BBPCV0456A. This action involves opening a 125VDC breaker (NK5108 for BBPCV0455A or NK4421 for BBPCV0456A). NK51 is located in fire area C-16 and NK44 is located in fire area C-15. C-15 and C-16 are the battery rooms for their respective trains.

Removal of DC power will cause the PORV to revert to its closed position. The recovery action was evaluated in the Task 11 Human Reliability Report (Callaway Fire PRA Report 17671-011).

Attachment B of that report provides the details of the calculation. A summary of the salient results is provided below. For all areas in which this action is credited, it was verified by walkdowns that a fire free path exists from the control room to the battery rooms. MAAP analysis was performed to determine the timing required to prevent core uncovery, assuming no ECCS was available. The assumption of no ECCS was used as a bounding case. Although one charging pump is guaranteed available in all sequences, the Boron Injection Header Isolation valves do not have a function to open in Nuclear Safety Capability Assessment (NSCA) and thus it is possible for some scenarios that they are both failed closed. The critical time for operator action is 1.83 hours9.606481e-4 days <br />0.0231 hours <br />1.372354e-4 weeks <br />3.15815e-5 months <br /> to prevent core uncovery.

The operator action is modeled as two basic eventsa cognitive error (OP-COG-FO-PORV) and an execution error (OP-OMA-FF-ISPORV). The HEPs assigned to each action are:

Page 119 of 136 to ULNRC-05876 Operator action HEP OP-COG-FO-PORV 9.4E-04 OP-OMA-FF-ISPORV 3.4E-05 TOTAL 9.75E-04 The operator actions that are credited in each fire area are considered within the context of all actions that may be required for a fire within that fire area. That is, the feasibility analysis, which is documented in the Nuclear Safety Capability Assessment calculation (KC-26), was performed by considering all actions that are required in each fire area. Then, the most limiting timeline from all of the applicable fire areas was selected as the basis for the final human error probability (HEP) calculation in the 17671-011 analysis.

Part 4: Risk Achievement Worth of the Recovery Action:

This RAI requested an evaluation of the effect of not crediting the recovery action in any scenario.

Tables 5 and 6 show that the results of not crediting the recovery action in any fire area is an increase in delta CDF of greater than 1.00E-7/yr. This increase was above the Callaway Plant NFPA 805 project guidelines for risk increase which led to the crediting of the action in several areas. The increase in fire risk if the local action to close the PORV is always failed was evaluated only for fire areas in which the recovery action is credited. These areas are indicated in Table 1, all other areas shown do not credit the recovery action, so it is already assumed to fail. To calculate the change in risk if the recovery action is always failed, the event OP-OMA-FF-ISPORV was set to fail (event value =

1.0) in the concatenated cutset equation files. These results are shown in Tables 5a and 5b.

Table 5a - Change in Risk with OP-OMA-FF-ISPORV = 1.0 (with CPT)

Baseline Risk ISPORV Failed Change CDF (per yr.) LERF (per yr.) CDF (per yr.) LERF (per yr.) CDF (per yr.) LERF (per yr.)

3.37E-06 1.95E-07 4.23E-06 2.01E-07 8.68E-07 6.70E-09 Table 5b - Change in Risk with OP-OMA-FF-ISPORV = 1.0 (no CPT)

Baseline Risk ISPORV Failed Change CDF (per yr.) LERF (per yr.) CDF (per yr.) LERF (per yr.) CDF (per yr.) LERF (per yr.)

4.77E-06 2.51E-07 5.67E-06 2.58E-07 9.01E-07 7.60E-09 As shown in the tables, removing all credit for the PORV LOCA recovery actions leads to CDF increases of approximately 9E-7/yr. and LERF increases of approximately 7E-9/yr. The increases are slightly larger when credit for CPTs is removed from the spurious actuation probability calculations.

All of the risk increase is delta risk, since the risk increases are due to removing credit for an NFPA 805 recovery action that is mitigating a VFDR. Since delta risk due to VFDRs is directly applicable to the Callaway Plant NFPA 805 application, these risk increases (which are equivalent to delta risk increases) are compared to plant-wide delta risk in Tables 6a and 6b.

Page 120 of 136 to ULNRC-05876 Table 6a - Risk Increases Compared to Plant-Wide Delta Risk (with CPT)

Plant-Wide ISPORV = 1.0 Theoretical Total CDF/CDF LERF/LERF CDF LERF Increase Increase CDF LERF (per yr.) (per yr.) (per yr.) (per yr.) (per yr.) (per yr.)

1.87E-06 3.84E-08 8.68E-07 6.70E-09 2.74E-06 4.51E-08 Table 6b - Risk Increases Compared to Plant-Wide Delta Risk (no CPT)

Plant-Wide ISPORV = 1.0 Theoretical Total CDF/CDF LERF/LERF CDF LERF Increase Increase CDF LERF (per yr.) (per yr.) (per yr.) (per yr.) (per yr.) (per yr.)

1.87E-06 3.84E-08 9.01E-07 7.60E-09 2.77E-06 4.60E-08 The differences between the results for the a and b case for delta risk is very small because the CPT credit was removed from both the variant condition and the compliant condition. The total delta risk in Tables 6a and 6b still have reasonable margin to the plant-wide delta risk limits of 1E-5/yr. and 1E-6/yr. for CDF and LERF, respectively.

Summary:

The Pressurizer PORV LOCA scenario has been shown to occur at a relatively low frequency in Table 2. A bounding assessment of the possible risk reduction if the PORV LOCA failure mode was completely eliminated at the Callaway Plant was shown in Tables 3a/b and 4a/b. Even if the PORV LOCA VFDR failure mode was completely eliminated from the Callaway Plant, the overall risk reduction would not affect the risk insights or conclusions of the Callaway Plant NFPA 805 Application Request. Most fire areas that have the PORV LOCA VFDR do not credit an NFPA 805 recovery action to reduce risk, as shown in Table 1. Table 5a/b shows the change in risk if that PORV LOCA recovery action credit was removed (i.e. if ISPORV is assumed to fail). Table 6a/b summarizes the change in risk if ISPORV is failed, and adds that risk increase to the plant-wide delta risk totals to create new "theoretical total" delta risk metrics. If ISPORV is assumed to fail in all fire areas, the resultant total plant delta risk is shown to maintain significant margin to the plant-wide delta risk limits of 1E-5/yr. and 1E-6/yr. for CDF and LERF, respectively.

Page 121 of 136 to ULNRC-05876 Probabilistic Risk Assessment RAI 14 In the LAR, reference is made to self approval in regard to the NFPA 805 transition results. This is incorrect as self approval thresholds do not apply for the transition aspects of the application, but rather are only applicable post-transition in evaluating plant change evaluations (i.e., at the time of the submittal the licensee has not been sanctioned by the NRC to self approve any fire-related plant changes). The licensee should revise this statement in connection with the transition risk results and indicate if revising this statement has any effect on the LAR.

Response to Probabilistic Risk Assessment RAI-14 Response provided by ULNRC-05851 dated April 17, 2012.

Probabilistic Risk Assessment RAI 15 Table W-1 of the LAR includes in its title "95% of Calculated Fire CDF," but the table only includes approximately 58 percent of the Fire CDF. Please provide clarification regarding the title and table and revise as appropriate.

Response to Probabilistic Risk Assessment RAI 15 Response provided by ULNRC-05851 dated April 17, 2012.

Page 122 of 136 to ULNRC-05876 Probabilistic Risk Assessment RAI 16 Scenarios 3801T3 and 3801T2 (upper cable spreading room transient fires) appear to involve fire-induced failure of Reactor Coolant Pump (RCP) seal cooling, Auxiliary Feedwater (AFW), and feed-and-bleed cooling resulting in core damage. The Conditional Core Damage Probability (CCDP) is stated as 0.76 indicating that some mitigative capability is available, albeit with a low probability of success. It is not clear to the NRC staff that these scenarios have acceptable defense-in-depth given the very high CCDP. During the audit, the licensee stated that the CCDP is artificially high due to calculation methods. Please discuss these scenarios in more detail including the mitigation capability that remains after fire damage and how that capability is consistent with adequate defense-in-depth. A discussion of key assumptions which impact these scenarios conservatively (if any) as well as administrative controls or other measures which reduce the likelihood of transient combustibles in the critical location should also be provided. A more accurate quantitative assessment of CCDP should be provided, or a justification as to why this is not possible.

Additional Justification Needed The second to last sentence in the 1st paragraph includes the sentence, During the FRE process, these conservatisms were discussed and the decision was made to retain the conservatisms in the fire PRA and attribute them to defense-in-depth. Conservatisms in the model are not generally defense-in-depth attributes. The sentence is not needed as the later discussion does address defense-in-depth.

Response to Probabilistic Risk Assessment RAI 16 These scenarios occur in room 3801, which is the Upper Cable Spreading room (Fire Area C-22).

Scenarios in the upper cable spreading room are subject to multiple analytical conservatisms that tend to increase the conditional risk calculation results. The total CDF for scenarios 3801T3 and 3801T2 is 7.62E-7/yr. The delta CDF is 9.06E-8/yr. While these two scenarios are the highest CDF contributors in Fire Area C-22, the absolute result is acceptable with respect to meeting the risk criteria in Regulatory Guides 1.205 and 1.174. There are several conservatisms inherent in these scenarios, which if removed would reduce the fire risk. During the FRE process, these conservatisms were discussed and the decision was made to retain the conservatisms in the fire PRA and attribute them to safety margin. These conservatisms are summarized below:

1st Conservatism: Additive Propagation of High Basic Event Probabilities for Seal Failure The dominant cutsets in this scenario involve loss of all RCP seal cooling due to spurious isolation of the seal injection lines and the Component Cooling Water [CCW] thermal barrier return line. The total quantified probability for loss of all RCP seal cooling is 3.1. This is due to the extraordinarily high probabilities postulated for spurious operation of valves. The top 4 cutsets for loss of seal cooling are 0.32 each. The next 8 cutsets are 0.16 each. The WinNUPRA code adds each cutset without regard to their exclusivity. The WinNUPRA code does not have a quantification capability to provide a min-cut-upper-bound result that adjusts for cutsets with high failure probabilities. If the numerical probability of the loss of seal cooling is capped at 1 (guaranteed failure) then the Conditional Core Damage Probability [CCDP] for the entire scenario is reduced to 0.25.

Page 123 of 136 to ULNRC-05876 2nd Conservatism Additive Propagation of High Basic Event Probabilities for SG Blowdown A similar effect occurs with the steam generator blowdown isolation valves failing open. Each valve is attributed a 0.8 probability of spurious opening. The AFW (decay heat removal) success criteria requires heat removal from two Steam Generators [SGs]. If each SG has a probability of failure of 0.8, then the total probability for Loss of AFW is 1.6. In this scenario, success could be achieved by feeding one SG with the A train Auxiliary Feedwater [AFW] motor driven pump and one SG with the Non-safety motor driven auxiliary feedwater pump. This success criteria combination is not modeled in the PRA analysis, but would be available in the event this fire scenario ever occurred.

Removing the calculation-related conservatisms, by altering the individual MOV spurious closure probabilities, the CCDP is reduced from ~0.76 to ~0.25. If the spurious valve actuation probabilities are changed to a logical 1 (i.e., guaranteed failure), the CCDP reduces to about 0.35, simply due to elimination of non-minimal cutsets. The point to be made on these scenarios is that there are several valves postulated to spuriously actuate with probabilities which exceed the bounds for rare-event approximation. The PRA codes show artificially high core damage probabilities when these high probabilities are input. The cutset results were reviewed and considered during the Fire Risk Evaluation [FRE] process and the decision was made to proceed with the over-estimation of CDF.

That does not mean the scenario has insufficient defense-in-depth or safety margin.

With respect to defense in depth for fire ignition, the upper cable spreading room will be administratively controlled as a No Transient Combustible Storage area and also as a No Hotwork area while the plant is at power. These controls, plus the absence of any fixed ignition sources, provide reasonable assurance that the likelihood of a challenging fire in the area is very low.

In summary, these upper cable spreading room fires have the following characteristics:

1. Low likelihood of a challenging fire, due to no fixed sources and administrative transient controls.

2. The area is provided with a wet-pipe suppression system.

3. Availability of the A Motor Driven Auxiliary Feedwater Pump [MDAFP] and the Non-Safety Auxiliary Feedwater Pump [NSAFP], which are individually failed by spuriously open SG Blowdown isolation valves, but the potential for success with both pumps in operation, is not considered.

4. Statistical over calculation of multiple events with high probabilities for spurious operation cause an over estimation of the CCDP.

These characteristics are considered to be sufficient justification that, even with high calculated conditional core damage probabilities; there is reasonable defense in depth available for these fire scenarios and the remaining mitigation capabilities provide an overall risk that is sufficiently low.

Page 124 of 136 to ULNRC-05876 Probabilistic Risk Assessment RAI 17 The change in risk or delta-risk (risk) for fire areas A-30 and TB-1 is identified as "0.00+00" while for fire area C-35 it is identified as "epsilon." Please clarify the intended difference between these table entries.

Additional Information Needed The response explicitly defines epsilon but not 0.00+00. The difference between the two remains unclear and seems to overlap. The response should explicitly state when 0.00+00 is used Response to Probabilistic Risk Assessment RAI 17 The table below shows the three situations for reporting delta risk. The (epsilon) is used when the fire ignition sources in the fire area do not cause any damage to Fire PRA components or cables.

Although the Fire PRA will not show any risk, a very small risk is possible such as a risk level below the truncation probability of the PRA. As such, these no fire-induced damage scenarios are considered to have very small risk and are denoted with the symbol .

DELTA RISK IS SHOWN AS SITUATION (from LAR Table W-2)

Fire Area has no VFDRs from NSCA (fire area is N/A deterministically compliant).

Fire Area contains VFDR(s), but detailed fire scenario modeling does not damage any Fire PRA cables or components.

Fire Area contains VFDR(s) and fire modeling shows fire damage to PRA cable(s) or component(s). Fire damage Calculated delta risk is shown, may or may not include damage to VFDR cables or which may be zero or non-zero components.

A zero delta risk is only used for two fire areas in Table W-2, A-30 (Auxiliary Feedwater Valve Compartment, SGs B & C) and TB-1 (Turbine Building). These fire areas have zero delta risk reported because the VFDRs assigned to these areas are not capable of being evaluated under the PRA success criteria. Thus, although NSCA identifies VFDRs, they are quantified in the PRA as having 0.00E+00 risk.

A-30 has a VFDR for the Steam Generator-C PORV spuriously opening (VFDR-A-30-001). SG-C is a non-credited steam generator in NSCA, but this is a VFDR because it causes RCS overcooling and interferes with pressurizer water level control. Fire PRA does not model pressurizer water level control, so this has no effect on the PRA.

The first two VFDRs for TB (TB-01-001 and TB-01-002), are related to cable damage to various Pressurizer heater power and control power supply cables. The FPRA does not model Pressurizer heaters because they are used to control and maintain sub-cooling in the RCS. The thermal-hydraulic Page 125 of 136 to ULNRC-05876 analyses upon which the PRA success criteria are based show that sub-cooling is not required to mitigate core damage during the postulated accident scenarios in the Internal Events and Fire PRA models. As such, the Pressurizer heaters are not included in the PRA models and the state of the heaters has no impact on calculated risk.

VFDR TB-01-003 is a loss of switchgear room B HVAC, which is not required in the FPRA for success. Even without HVAC for the duration of the FPRA mission time, room temperatures remain low enough to allow unaffected performance of components in the Switchgear Rooms. As such, a loss of Switchgear Room HVAC does not impact the plants ability to mitigate an accident.

Probabilistic Risk Assessment RAI 18 The disposition of VFDRs with regards to defense-in-depth and safety margins in the LAR in Attachment C provides no technical justifications but simply states an evaluation was performed and found to be acceptable. Please describe the process that was applied to evaluate the acceptability of defense-in-depth and safety margins for VFDRs. (This should be a general description of the process and criteria, not a detailed basis for each VFDR.) The description should also address how reliance upon multiple, time-critical, or complex recovery actions for a particular fire scenario is evaluated to assure there is no over reliance upon operator actions, and how the risk evaluations for recovery action probabilities consider multiple actions in a single scenario.

Response to Probabilistic Risk Assessment RAI 18 Response provided by ULNRC-05851 dated April 17, 2012.

Page 126 of 136 to ULNRC-05876 Probabilistic Risk Assessment RAI 19 There is no description of how the change-in-risk is estimated for the various VFDRs. Please provide a description about the modeling of the cause-and-effect relationship in the PRA for each type of VFDR (e.g., cable separation issues, degraded barriers). The description should also include any key assumptions or conservatisms in these evaluations including, for example, if recovery actions are included.

Additional Information Needed When calculating delta risk between cases (b) and (c), i.e., with vs. without credit for an operator action, where (c) is the ideal compliant case (lowest risk), does the nominal HEP in case (b) include fire effects (and is therefore higher), leading to a greater delta risk?

Response to Probabilistic Risk Assessment RAI 19 Preparing For Delta Risk Calculation VFDRs are identified by the NSCA, based on complete burnout of a FIRE AREA. A VFDR exists in the AREA if any NSCA safety function cannot be provided because sufficient trains of a mitigating system are failed by the fire. These VFDRs are then assessed by the PRA for calculation of delta risk.

The first step is for the PRA to determine if the NSCA safety criteria are reflected in the PRA. Safety functions specified in the NSCA analysis, but not included in the PRA analysis, have not been assessed for variant risk. These typically involve, for example, safety functions associated with maintaining pressurizer water level and RCS subcooling. The PRA does not require maintenance of pressurizer water level or subcooling so these VFDRs do not contribute to risk because they do not impact functions modeled in the PRA.

The next step is to determine if the VFDR components are explicitly modeled in the PRA. An example of components not explicitly modeled in the PRA would be instrumentation required for operator actions. For these VFDRs, it is necessary to identify a surrogate component in the PRA, which can be varied to assess the delta risk. For instruments, the surrogate basic event is generally the human error which requires the instrument as a cue. Alternatively, it could be the component to be actuated by the operator.

If a VFDR matches a PRA success criteria and credited system, it is then quantitatively analyzed in the Fire Risk Evaluation process. In these cases, the disposition of the VFDR is based on a delta risk calculation. The delta risk is calculated as the difference between the variant case (which is the proposed configuration of the post-transition Callaway plant) compared to the compliant case, which is a hypothetical configuration in which the Callaway plant complies with all deterministic requirements of NFPA-805. This hypothetical, compliant configuration is referred to as the deterministic compliant case in the Callaway Fire Risk Evaluations (FRE) and Fire Safety Analysis Page 127 of 136 to ULNRC-05876 (FSA) reports.

The majority of VFDRs evaluated in the PRA are cable separation issues. The cable separation VFDRs occur because cables or equipment for both trains of a mitigating system are located in the same Fire Area. Another category of VFDRs involves cable or equipment damage which can put the plant outside of its NSCA mitigation capabilities. These include spurious closure of a RCP seal return isolation valve or spurious opening of a steam generator atmospheric relief valve. These will be VFDRs regardless of where the cables are located.

Another type of VFDR is related to degraded fire wrap. In these cases, the risk associated with the VFDR is treated by not giving credit to the degraded wrap in the fire modeling and associated input to the Fire PRA. That is, the cables within the degraded wrap were treated as though there was no wrap present. Like the cable separation VFDRs, the basic event(s) corresponding to the VFDR cable(s) are then set to zero to simulate compliance in the deterministically compliant risk calculation.

A final type of VFDR at the Callaway plant is the HDPE piping present in Fire Area C-1, which is potentially vulnerable to fire damage. To treat the risk associated with this VFDR, the HDPE pipe was treated as a potential target in the fire modeling. If the piping was failed in a postulated fire scenario, the deterministically compliant risk calculation set the basic events associated with that damage state to zero. Overall, the HDPE piping type of VFDR was treated the same as a cable separation issue.

Calculation Of Delta Risk All VFDRs are analyzed for delta risk with the same process. This process is summarized below:

The components associated with the VFDR are identified and the PRA basic events associated with the component failure modes are identified. Recovery actions associated with the VFDR are identified and verified to be in the model associated with the variant component. Three risk cases can be calculated, depending on the magnitude of the fire risk:

a) The VFDR is modeled as it is in the current plant configuration with the recovery action set to fail - i.e., this is the current risk with no recovery action.

b) The VFDR modeled with the recovery action at its nominal human error probability, which includes fire effects - i.e., this is the post-transition case.

c) The plant without the VFDR - i.e., assume the offending cables are separated or wrapped, or otherwise not failing. VFDRs are eliminated from the PRA model by setting the basic event(s) that are associated with the offending cable(s) to zero to simulate compliance in the deterministically compliant risk calculation.

The difference in risk between case a & c indicates if the VFDR is significant enough to require a recovery action. The difference in risk between case b and case c indicates if the recovery action provides an acceptable risk mitigation strategy. If the delta risk between case b and case c (cumulative for all VFDRs in the plant) does not meet RG 1.205 risk criteria, a plant change is required.

Page 128 of 136 to ULNRC-05876 Recovery actions for VFDRs are referred to as 805 Recovery Actions in the Callaway LAR, are not necessarily credited for all VFDRs. If the delta risk between case a & c is small enough, it is acceptable to leave the VFDR in the plant design without a mitigating recovery action.

VFDRs are identified by NSCA on the basis of complete burnout of an entire FIRE AREA. If the fire area is modeled in the PRA as whole room burnout, then the VFDRs are evaluated in a consistent basis with the NSCA. For areas which have fire modeling, the VFDRs are evaluated on the basis of each fire scenario. Whereas an area may have a VFDR in which two cables are located on opposite sides of the room, when fire modeling is employed, the two cables do not appear in the same scenario, so from the perspective of the PRA, the VFDR does not exist. Or, it is said the VFDR has no quantifiable risk, because it does not appear in any scenario.

Probabilistic Risk Assessment RAI 20 Please provide confirmation that the use of the guidance from EPRI TR-1016735, "Fire PRA Methods Enhancements, Additions, Clarifications, and Refinements to EPRI 1011989," included any modifications of this report as incorporated into Supplement 1 of NUREG/CR-6850.

Response to Probabilistic Risk Assessment RAI 20 Response provided by ULNRC-05851 dated April 17, 2012.

Probabilistic Risk Assessment RAI 21 The disposition for F&O PRM-B4-1 states that the fire-induced risk model report was updated to provide the bases for fire-induced initiators and non-applicability of Supporting Requirement PRM-B4. Please provide these bases.

Response to Probabilistic Risk Assessment RAI 21 Response provided by ULNRC-05851 dated April 17, 2012.

Page 129 of 136 to ULNRC-05876 Probabilistic Risk Assessment RAI 22 The dispositions of F&Os FSS-E03-1 and UNC-A1-2, both related to Supporting Requirement FSS-E3, cite conservatism in method selection and use of data from NUREG/CR-6850 as justification for not meeting the requirement (at Capability Category II) to provide a mean value and statistical representation of uncertainty intervals for parameters used to model significant fire scenarios. Please explain how the requirements of FSS-E3 are met or justify why they need not be.

Additional Information Needed:

The statement that there is no SR to incorporate the fire modeling uncertainties into the CDF/LERF equation uncertainty is misleading, and the conclusion that CC-II is attained questionable (with respect to SR FSS-E3; SR UNC-A1, to which this RAI also applies, is either Met or Not Met). As per SR UNC-A1, an uncertainty analysis in accordance with HLR-QU-E and its SRs in Part 2 must be performed. SR QU-E3 in Part 2 requires an estimation of the uncertainty interval of the CDF results consistent with the characterization of parameter uncertainties for CC-I, with the additional requirement to take into account the state-of-knowledge correlation for CC-II. Thus, it is not sufficient to just discuss the uncertainties on the individual parameters that contribute to CDF, but at least an estimate of the uncertainty on CDF itself is required. Provide this estimate. Note also that the term fire modeling as used here is not restricted to fire modeling just in the phenomenological sense, such as empirical correlations, zone models, of CDF models, but applies to all the elements that are input into the fire risk equation (including ignition frequency, non-suppression probability, etc.). (Note: a roll-up of the various sensitivity analyses performed in response to other PRA RAIs may form a reasonable estimate of the CDF uncertainty interval.)

Response to Probabilistic Risk Assessment RAI 22 Supporting Requirement FSS-E03 of the ASME PRA standard requires identification of numerical uncertainty bounds for fire modeling parameters in order to meet Capability Category II. Capability Category I allows qualitative assessment of uncertainty. The F&O indicates the Callaway Plant Fire PRA has met Capability Category I, but has not provided numerical uncertainty bounds for fire modeling parameters and thus does not met Capability Category II.

Several sources of uncertainty were considered in the fire Modeling. These are discussed quantitatively and qualitatively through the documented reports of the Callaway Plant NFPA 805 project. These are discussed in response to Fire Modeling RAI-2a and are equally applicable to this RAI.

The discussion and identification of uncertainty bounds is sufficient to attain Capability Category II.

There is no SR to incorporate the fire modeling uncertainties into the CDF/LERF equation uncertainty.

To provide an estimate of the Fire CDF and LERF, the various sensitivities that have been performed during the creation of the LAR and the subsequent RAI responses were summarized and compared.

The sensitivity studies and their results are shown in the tables below.

Page 130 of 136 to ULNRC-05876 Table 1 - Summary of Fire PRA Sensitivity Studies

1. Sensitivity Description Ignition source bins with > 1.0 use 6850 mean frequencies 1 FAQ 08-0048 instead of FAQ 08-0048 (also discussed in PRA RAI 07f)

Examine the impact of lower damage threshold and self-ignited 2 Thermoplastic cable fires for non-IEEE-383 cable in the Turbine Building (also discussed in PRA RAI 07f) 3 PRA RAI 07c Sets certain HEPs with relatively low time window margins to 1.0 4 PRA RAI 08b Removes the reduction factor from bin 16.2 Re-calculates plant-wide risk and delta risk without credit for 5 PRA RAI 09a CPTs Uses the 95th% ignition frequencies from NUREG/CR-6850 6 PRA RAI 09b Supplement 1 for all ignition source bins to study risk sensitivity to ignition frequency Considers the effect of both removing the PORV LOCA VFDR 7 PRA RAI 13 failure mode and assuming that the applicable recovery action always fails Table 2 - Fire PRA Sensitivity Quantitative Results Summary CDF LERF CDF LERF

1. Increase Increase Increase Increase (per yr.) (per yr.) (per yr.) (per yr.)

11 3.64E-06 7.77E-08 1.95E-06 4.50E-08 2 8.81E-07 2.29E-08 0.00E+00 0.00E+00 3 3.72E-07 7.91E-09 3.72E-07 7.91E-09 4 8.18E-09 1.59E-10 0.00E+00 0.00E+00 5 6.26E-06 1.48E-07 1.75E-06 7.36E-08 6 4.48E-05 8.34E-07 3.96E-06 7.96E-08 7 7.49E-07 3.53E-09 7.49E-07 3.53E-09 Total 5.31E-05 1.02E-06 6.83E-06 1.65E-07 Increases Plant 2.02E-05 3.97E-07 1.87E-06 3.84E-08 Totals Theoretical 7.33E-05 1.41E-06 8.70E-06 2.03E-07 Totals Risk Goals < 1E-4 < 1E-5 < 1E-5 < 1E-6 Results satisfactory satisfactory satisfactory satisfactory Note 1, sensitivity #1 is considered to be a subset of #6 and, as such, is not included in the sum for the Total Increase row.

Page 131 of 136 to ULNRC-05876 Each of these sensitivity studies generally contains conservatisms, as discussed in their corresponding RAIs and/or FPRA report sections. As such, combining all of the risk increases from these independent sensitivity studies cascades the conservatisms to provide a truly conservative, bounding estimate of fire risk uncertainty.

As noted in Table 2, even if all of the sensitivity study risk increases are summed and compared to the risk goals from Regulatory Guide 1.205, the totals are below the regulatory limits. As described above, these total risk metrics are subject to considerable conservatism and can be considered bounding for fire risk uncertainty.

Page 132 of 136 to ULNRC-05876 Probabilistic Risk Assessment RAI 23 In some of the Fire Evaluation of Delta Risk Calculations for the various fire areas, a HRR profile for a transient combustible less than the recommended (142 kW and 317 kW at the 75th and 98th percentiles) was assumed. Please provide the bases for these assumptions.

Additional Information Needed When citing as a basis for limiting the transient combustible HRR to 69 kW (98th %ile) that 69 kW is judged to be no larger than the 75th %ile fire in an electrical cabinet with qualified cable, further note, if correct, that this assumes no more than one cable bundle involved in the fire. Table E-1 of NUREG/CR-6850 reports two 75th %ile values for qualified cable, one at 69 kW (single bundle), the other at 211 kW (multiple bundles).

In the response to PRA RAI 23, the following needs additional clarification. The last bullet in the list of the basis for assuming a peak transient HRR of 69 kW cites the test results from Table G-7 of NUREG/CR-6850, stating that ... the types of fires that can be expected in these rooms (i.e.,

polyethylene trash can or bucket containing rags and paper) were measured at peak HRRs of 34 kW or below. This is true except for one test, SNL-Nowlen, Test #9, where a 50 kW HRR during the first 15 min was observed. Since this test involved a much larger trash can (30 gal) than any of the others (maximum of 5 gal), the bullet should add either of the following clarifications: (1) i.e., a maximum-sized polyethylene trash can or bucket of ~5 gal. containing rags or paper; or (2) were measured at peak HRRs of 50 kW or less. Provide the appropriate clarification.

Response to Probabilistic Risk Assessment RAI 23 EPRI-led Fire PRA Methods Review Panel issued decisions on methods submitted for their review.

Letter from NEI to NRC, B. Bradley to D. Harrison, Recent Fire PRA Methods Review Panel Decisions: Clarifications for Transient Fires and Alignment for Pump Oil Fires dated September 27, 2011 provided as a clarification to the guidance of NUREG/CR-6850 and part of the new PRA methods. Attachment 1 Description of Treatment for Transient Fires, and Attachment 3 Panel Decision, allow the user to choose a lower screening heat release rate for transient fires in a fire compartment based on the specific attributes and considerations applicable to that location. The guidance indicates that plant administrative controls should be considered in the appropriate HRR for a postulated transient fire and that a lower screening HRR can be used for individual plant specific locations if the 317kW value is judged to be unrealistic given the specific attributes and considerations applicable to that location.

At Callaway Plant, a 69kW transient heat release rate was justified for certain fire areas based on several factors:

Page 133 of 136 to ULNRC-05876

• All fire areas which were credited for a reduced heat release rate are subject to strict combustible controls (areas designated as No Storage) and so paper, cardboard, scrap wood, rags and other trash shall not be allowed to accumulate in the area.

• Large combustible liquid fires are not expected in these areas 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 these fire areas.

• 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 of NUREG/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 of NUREG/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.

Since only small quantities of trash in temporary containers can be expected, a 69kW 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.

Page 134 of 136 to ULNRC-05876 Section 6: Licensee Identified Changes (LIC)

LIC-01 Provided by ULNRC-05851 dated April 17, 2012 LIC-02 Provided by ULNRC-05851 dated April 17, 2012 LIC-03 Provided by ULNRC-05851 dated April 17, 2012 LIC-04 Provided by ULNRC-05851 dated April 17, 2012 LIC-05 Provided by ULNRC-05851 dated April 17, 2012 LIC-06 Provided by ULNRC-05851 dated April 17, 2012 LIC-07 Provided by ULNRC-05851 dated April 17, 2012 LIC-08 Provided by ULNRC-05851 dated April 17, 2012 LIC-09 LAR Table 4-3 and Attachment C have been revised to report Fire Area TB-1 Room 3307 fire suppression system as Y* under the Required for Chapter 4 Separation Criteria column. The note in Table 4-3 correctly indicates the system is required for Chapter 3 compliance.

LIC-10 LAR Attachment J was revised to include positive statements that the models are used within their validated range. Section 4.5.1.2 of the LAR has been updated to provide additional discussion related to the statements made in Attachment J.

LIC-11 LAR Table 4-3 and Attachment C have been revised to indicate that fire detection in Fire Zones 1206 and 1207 and Detection Zone 100 in Fire Zone 1130 are required to support NFPA 805 Chapter 4 Separation requirements.

LIC-12 LAR Table 4-3 and Attachment C have been revised to remove duplicate fire zones reported in Fire Area TB-1, Fire Zones 3706 and 3612. The entry for Fire Zone 3706 designated as "Fire Brigade Storage Area" has been removed. The entry for Fire Zone 3612 designated as "Field Office" has been removed.

LIC-13 It was not clear to the staff what the deficiency in the fire PRA model is relative to F&O PRM-B2-1 nor if the item was resolved by making changes or by simply clarifying the underlying issue. Changes have been made in Attachment V to address this.

LIC-14 Attachment V has been revised to clarify the basis for closing F&O PRM-B4-1.

LIC-15 The NRC has identified that the use of the Electric Power Research Institute (EPRI)

Technical Report TR-1006756 constitutes use of performance based methods for NFPA 805 Chapter 3 requirements. Therefore, a new NRC Approval Request 3 for use of the EPRI document has been added to Attachment L.

LIC-16 The LAR Table B-1 Section 3.4.1(a) is revised and a new LAR/TR Attachment T, Clarification of Prior NRC Approval Request 6, has been added to address the 2 hr grace period for Fire Brigade staffing allowed during Operations Main Control Room shift change. The allowance for the 2 hr grace period was originally included in the Westinghouse Standardized Technical Specifications that were approved for Callaway Plant.

LIC-17 The LAR Table B-2 Sections 3.5.1.5 and 3.5.2.3 are revised to provide more detail regarding the Callaway Plant method for evaluation of spurious actuations for ungrounded DC circuits.

Page 135 of 136 to ULNRC-05876 LIC-18 Ameren Missouri has revised LAR Table B-3 and Table 4-3 for fire areas A-1, A-16, A-27, and C-1 to identify that a 20 ft. separation zone is credited as allowed by NFPA 805 Section 4.2.2.3(b) within the fire area to meet deterministic requirements. This information is added to the Fire Area Comments section of the B-3 table and to the features section of Table 4-3. Additionally, LAR Attachment X, Other Requests for Approval, has been revised to add a new request for NRC approval of the 20 ft.

separation zones in fire areas A-1, A-16, A-27 and C-1. Additionally, the Fire Safety Analysis for each fire area has been updated to enhance the descriptions of the 20 ft.

separation zones to be consistent with the level of detail found in Attachment X. Also, LAR Table B-3 and Table 4-3 for fire area RB-1 have been updated to credit the 20 ft.

separation zones previously included in LAR Attachment X.

LIC-19 NFPA 805, Section 2.7.3 contains the requirements for Quality Assurance and in response; LAR / TR Section 4.7.3 provided a description of the Post Transition QA Program. This LIC revises that description of the Post Transition QA Program specifically discussing the FP QA Program requirement to conduct independent audits of the FP Program by the Nuclear Oversight Department. The revised text results in two changes: 1) the location of the description of the audits which includes the scope is being removed from the Operating Quality Assurance Manual (OQAM) and placed in FSAR Standard Plant (SP) Section 9.5.1 with the FP Program QA requirements such that all FP QA Program requirements will be consolidated within that section of the FSAR, and 2) the frequency at which the audits must be conducted as stated in OQAM Section 18 is being revised from 2 years to 3 years. There is no change to the scope of the 3 year audit; it remains the same as stated in the current QA program. Additionally, there is no change to the requirements for personnel conducting the audits.

Page 136 of 136

Attachment A: Revisions to Transition Report Attachment A - NEI 04-02 Table B Transition of Fundamental Fire Protection Program and Design Elements

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.2.3(1) Inspection, testing, and maintenance for Complies, with Required See implementation items Procedure APA-ZZ-00703, "Fire fire protection systems and features Action identified below. Protection Operability Criteria and credited by the fire protection program Surveillance Requirements," Rev. 20 /

All Calculation KC-162, "Performance Based Fire Protection Surveillance Frequency Program," Rev. 0 Procedure APA-ZZ-00700, "Fire Protection Program," Rev. 18 CAR 201101832, "Track Implementation Items for NFPA-805 Project" / All IMPLEMENTATION ITEMS:

11-805-048 Procedures APA-ZZ-00700 and APA-ZZ-00703 will be revised to include inspection, testing, and maintenance requirements for all fire protection systems and features credited by the fire protection program.

11-805-069 During the implementation of the NFPA 805 license basis, performance-based surveillance frequencies will be established as described in Electric Power Research Institute (EPRI) Technical Report TR-1006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features" and evaluated in Callaway Plant Calculation KC-162, "Performance Based Fire Protection Surveillance Frequency Program."

Inspection, testing, and maintenance for Submit for NRC NRC approval of the use of EPRI None fire protection systems and features Approval Technical Report TR-1006756 in credited by the fire protection program establishing performance-based inspection, testing, and LIC maintenance frequencies for fire 15 protection systems and features credited by the fire protection program is being requested in Attachment L.

August 2011 Page A-5

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.3.7.1 Storage of flammable gas shall be Complies, with Required Bulk hydrogen complies with the Calculation KC-27, "NFPA Code located outdoors, or in separate detached Action requirements of NFPA 50A-1973 Conformance Review," Rev. 0 / FPE buildings, so that a fire or explosion will Edition. Exceptions requiring Appendix A, Section 50A RAI not adversely impact systems, further action are identified below. 01 equipment, or components important to NFPA 50A, "Standard for Gaseous nuclear safety. NFPA 50A, Standard for Hydrogen Systems at Consumer Sites,"

Gaseous Hydrogen Systems at 1973 Edition / All Consumer Sites, shall be followed for hydrogen storage. CAR 201101832, "Track Implementation Items for NFPA 805-Project" / All IMPLEMENTATION ITEMS:

07-050A-001 Procedures will be revised to ensure that the hydrogen supply system is inspected annually and maintained by Ameren Missouri.

07-050A-002 Dry vegetation and combustible material within 15 feet of the hydrogen supply area will be removed. Additionally, procedures will be revised to ensure that the area within 15 feet of the hydrogen supply area is kept free of dry vegetation and combustible materials.

3.3.7.2 Outdoor high-pressure flammable gas Complies No Additional Clarification FSAR Site Addendum (SA), Rev. OL-storage containers shall be located so 15 / Section 2.2.2.1.2.1 that the long axis is not pointed at buildings.

3.3.7.3 Flammable gas storage cylinders not Complies No Additional Clarification Safe Work Practices Manual, Rev. 18 /

required for normal operation shall be "Compressed Gases" Section isolated from the system.

August 2011 Page A-25

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.4.1(a) On-Site Fire- A fully staffed, trained, and equipped fire- Complies by Previous Per Section 6.2.2.e of NUREG- NUREG-0830, "Safety Evaluation Fighting Capability. fighting force shall be available at all NRC Approval 1058, "Technical Specifications Report Related to the Operation of times to control and extinguish all fires on Callaway Plant, Unit No. 1," "A site Callaway Plant, Unit No. 1," dated site. This force shall have a minimum Fire Brigade of at least five October 1981 / Section 9.5.1.6 complement of five persons on duty and members (may be less than the shall conform with the following NFPA minimum requirements for a period NUREG-1058, "Technical LIC standards as applicable: of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in Specifications Callaway Plant, Unit No. 16 order to accommodate unexpected 1" / Section 6.2.2.e absence provided immediate action is taken to fill the required positions) shall be maintained onsite at all times."

Per Section 9.5.1.6 of NUREG-0830, entitled "Administrative Controls, Fire Brigade, Technical Specifications, and Training," "The applicant has committed to follow the staff Standard Technical Specifications. The staff finds this acceptable."

August 2011 Page A-34

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.4.1(a)(1) NFPA 600, Standard on Industrial Fire Complies, with Required The industrial fire brigade complies Calculation KC-27, "NFPA Code Brigades (interior structural fire fighting) Action with NFPA 600-2000 Edition. Conformance Review," Rev. 0 /

Exceptions requiring further action Appendix A, Section 600 FPE are identified below. RAI NFPA 600, "Standard on Industrial Fire 02 Brigades," 2000 Edition / All CAR 201101832, "Track Implementation Items for NFPA-805 Project" / All IMPLEMENTATION ITEMS:

07-600-001 A safety and health policy will be documented for the Callaway Plant Fire Brigade. The policy will satisfy the requirements of NFPA 600, Sections 2-1.4 and 2-2.4.

07-600-002 Fire brigade policy documents and procedures will be updated to include a requirement for a standard system to identify and account for each industrial fire brigade member present at the scene of the emergency, in accordance with NFPA 600, Section 2-2.1.4.

The requirement will also meet NFPA 600, section 2-4.5, and will specify that industrial fire brigade members be issued identification for the following purposes:

(1) Assistance in reaching the incident in an emergency (2) Identification by security personnel (3) Establishing authority 07-600-003 A risk management policy will be written for emergency response. The risk management policy shall be routinely reviewed with industrial fire brigade members and shall be based on the following recognized principles:

(1) Some risk to the safety of industrial fire brigade members is acceptable where saving human lives is possible.

(2) Minimal risk to the safety of the industrial fire brigade members, and only in a calculated manner, is acceptable where saving endangered property is possible.

(3) No risk to the safety of industrial fire brigade members is acceptable where saving lives or property is not possible.

07-600-004 The Callaway Plant Fire Brigade training program will be updated to include a periodic review of NFPA 600.

FPE 07-805-015 A requirement that specifies that fire brigade protective clothing and respiratory protective equipment shall conform to the applicable NFPA standard will be documented in APA-ZZ-00700.

RAI 02 August 2011 Page A-34

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.4.1(e) Each industrial fire brigade member shall Complies No Additional Clarification Procedure APA-ZZ-00912, "Callaway pass an annual physical examination to Plant Medical Physical Program," Rev.

determine that he or she can perform the 16 / Section 4.3 strenuous activity required during manual firefighting operations. The physical Procedure APA-ZZ-01000, "Callaway examination shall determine the ability of Radiation Protection Program" (CTSN each member to use respiratory 4111), Rev. 33 / Section 4.18 protection equipment.

3.4.2 Pre-Fire Plans. Current and detailed pre-fire plans shall Complies, with Required See implementation item identified Callaway Plant Fire Preplan Manual, FPE be available to the industrial fire brigade Action below. Rev. 34 / All RAI for all areas in which a fire could jeopardize the ability to meet the CAR 201101832, "Track 03 performance criteria described in Section Implementation Items for NFPA-805 1.5. Project" / All IMPLEMENTATION ITEMS:

11-805-076 The Fire Pre-Plan Manual will be revised as follows:

• The fire pre-plan attachments will be revised where the radiation release criteria are applicable for gaseous and liquid effluent as described in Table E-1/E-2 to include effluent controls and monitoring.

• New Pre-Fire Plans will be added for C-36 and C-37.

• Two new Attachments will be added, for Temporary Structures Inside the PA and for Temporary Structures Outside the PA, and existing Fire Attack Guidelines will be combined into each attachment.

August 2011 Page A-37

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.4.2.2 Pre-fire plans shall be reviewed and Complies No Additional Clarification Procedure APA-ZZ-00700, Fire updated as necessary. Protection Program, Rev. 18 / Section 3.4.8 3.4.2.3 Pre-fire plans shall be available in the Complies, with Required See implementation item identified Procedure APA-ZZ-00700, Fire control room and made available to the Action below. Protection Program, Rev. 18 / Section plant industrial fire brigade. 3.4.8 CAR 201101832, "Track Implementation Items for NFPA-805 Project" / All IMPLEMENTATION ITEMS:

FPE 07-805-047 A statement will be added to procedure APA-ZZ-00700 to require that controlled copies of the pre-fire plans be maintained in the RAI Control Room and made available to the fire brigade. 04 3.4.2.4 Pre-fire plans shall address coordination Complies with The pre-fire plans do not address Procedure OTO-KC-00001, "Fire with other plant groups during fire Clarification coordination with other plant Response," Rev. 8 / Step 15 emergencies. groups, this information is FPE contained within the referenced Procedure EIP-ZZ-00226, "Fire RAI procedures, which are used in Response Procedure for Callaway 04 conjunction with the pre-fire plans Plant," Rev. 14 / Section 5.2 as part of the overall fire response.

3.4.3 Training and Industrial fire brigade members and other N/A N/A - General statement; No N/A Drills. plant personnel who would respond to a technical requirements fire in conjunction with the brigade shall be provided with training commensurate with their emergency responsibilities.

August 2011 Page A-39

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.4.4 Fire-Fighting Protective clothing, respiratory protective Complies, with Required Equipment is provided for the fire Procedure APA-ZZ-00700, "Fire equipment. equipment, radiation monitoring Action brigade as required. Per visual Protection Program," Rev. 18 / All equipment, personal dosimeters, and fire inspection of equipment, it is in suppression equipment such as hoses, accordance with applicable NFPA Procedure APA-ZZ-00743, "Fire Team nozzles, fire extinguishers, and other codes, as documented in CAR Organization and Duties," Rev. 23 / FPE needed equipment shall be provided for 200902315. See implementation Section 4.1.3.e RAI the industrial fire brigade. This equipment item identified below. 05 shall conform with the applicable NFPA Procedure HTP-ZZ-05006, "Fire standards. Involving Radioactive Material or Entry into the Radiologically Controlled Area," Rev. 9 / Section 6.1.2 HDP-ZZ-08000, "Respiratory Protection Program," Rev. 21 / Section 3.9.2 Calculation KC-27, "NFPA Code Conformance Review," Rev. 0 /

Appendix A, Section 600 CAR 200902315, "NFPA 805 Transition - Site Organizations Support Tracking CAR" / All CAR 201101832, "Track Implementation Items for NFPA-805 Project" / All Procedure APA-ZZ-00700, "Fire Protection Program," Rev. 18 / All IMPLEMENTATION ITEMS:

07-805-015 A requirement that specifies that fire brigade protective clothing and respiratory protective equipment shall conform to the applicable NFPA standard will be documented in APA-ZZ-00700.

August 2011 Page A-44

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.5 Water Supply N/A N/A N/A - General statement; No N/A technical requirements 3.5.1 A fire protection water supply of adequate N/A N/A - General statement; No N/A reliability, quantity, and duration shall be technical requirements provided by one of the two following methods.

FPE 3.5.1(a) Provide a fire protection water supply of N/A Callaway Plant complies with N/A RAI not less than two separate 300,000-gal subsection (b) to this requirement; (1,135,500-L) supplies. therefore, compliance with 11 subsection (a) is not required.

August 2011 Page A-46

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.5.1(b) Calculate the fire flow rate for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Complies with Per the references identified, the Calculation M-650-00071, "Hydraulic This fire flow rate shall be based on 500 Clarification largest design demand of any Calculations for Turbine Building EL gpm (1892.5 L/min) for manual hose credited sprinkler or fixed water 2000-0 South, Standardized Nuclear streams plus the largest design demand spray system in the power block is Unit Power Plant System - SNUPPS FPE of any sprinkler or fixed water spray SKC29 at 2300 gpm. The fire flow 10466-M-650," Rev. 1 / All RAI system(s) in the power block as rate is 2300 gpm + 500 gpm hose 11 determined in accordance with NFPA 13, stream allowance = 2800 gpm. Calculation M-KC-316, "Fire Protection Standard for the Installation of Sprinkler The total amount of water flowed System Hydraulic Calculations Systems, or NFPA 15, Standard for over two hours would be 2800 gpm Determine the Adequacy of the Fire Water Spray Fixed Systems for Fire x 120 min = 336,000 gallons. Per Protection System for Providing the Protection. The fire water supply shall be the references identified, an Design Flow and Pressure to the capable of delivering this design demand adequate reliability, quantity, and Interface with the Sprinkler System,"

with the hydraulically least demanding duration is available to meet this Rev. 1C / All portion of fire main loop out of service. demand.

Calculation M-KC-413, "Fire Protection Determines the Flow Requirements of the Fire Pump," Rev. 0 / All Drawing F/P 095067, "Fire Protection System Fire Water Storage Tank General Plan," Rev. 4 / All Procedure APA-ZZ-00703, "Fire Protection Operability Criteria and Surveillance Requirements," Rev. 20 /

Section 4.3.3.a.1 August 2011 Page A-47

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.5.2 The tanks shall be interconnected such Complies No Additional Clarification Calculation KC-27, "NFPA Code that fire pumps can take suction from Conformance Review," Rev. 0 /

either or both. A failure in one tank or its Appendix A, Section 22 piping shall not allow both tanks to drain.

The tanks shall be designed in Drawing F/P 095067, "Fire Protection LIC-accordance with NFPA 22, Standard for System Fire Water Storage Tank 01 Water Tanks for Private Fire Protection. General Plan," Rev. 4 / All Exception No. 1: Water storage tanks NFPA 22, "Standard for Water Tanks shall not be required when fire pumps are for Private Fire Protection," 1974 able to take suction from a large body of Edition / All water (such as a lake), provided each fire pump has its own suction and both suctions and pumps are adequately separated.

Exception No. 2: Cooling tower basins shall be an acceptable water source for fire pumps when the volume is sufficient for both purposes and water quality is consistent with the demands of the fire service.

August 2011 Page A-48

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.5.3 Fire pumps, designed and installed in Complies No Additional Clarification Calculation M-KC-316, "Fire Protection LIC-accordance with NFPA 20, Standard for System Hydraulic Calculations 02 the Installation of Stationary Pumps for Determine th Adequacy of the Fire Fire Protection, shall be provided to Protection System for Providing the ensure that 100 percent of the required Design Flow and Pressure to the flow rate and pressure are available Interface with the Sprinkler System,"

assuming failure of the largest pump or Rev. 1C / All pump power source.

Calculation M-KC-413, "Fire Protection Determines the Flow Requirements of the Fire Pump," Rev. 0 / All Calculation KC-27, "NFPA Code Conformance Review," Rev. 0 /

Appendix A, Section 20 NFPA 20, "Standard for the Installation of Centrifugal Fire Pumps," 1974 Edition / All LIC-03 and LIC-05 August 2011 Page A-49

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.5.4 At least one diesel engine-driven fire Complies by Previous Per Section 9.5.1.1 of NUREG- NUREG-0830, "Safety Evaluation pump or two more seismic Category I NRC Approval 0830, "The water supply system Report Related to the Operation of Class 1E electric motor-driven fire pumps consists of three fire pumps Callaway Plant, Unit No. 1," dated connected to redundant Class 1E separately connected to a buried, October 1981 / Section 9.5.1.1 emergency power buses capable of 14-in pipe loop around the plant.

providing 100 percent of the required flow There are three 50-percent Letter ULNRC-00189 from Bryan (UE) rate and pressure shall be provided. capacity fire pumps, each rated at to Rusche (NRC) dated April 15, 1977 1500 gpm at 347-ft head. One / Section 9.5.1.1 pump is electric motor driven and LIC-two are diesel engine driven."

04 "Based on this evaluation, the staff concludes that the water supply system is adequate, meets the guidelines of Section E.2 of Appendix A to BTP ASB 9.5-1. and is, therefore, acceptable."

The fire pump configuration, as approved in the referenced SER, is still in the same configuration as that which was approved. There have been no plant modifications or other changes that would invalidate the basis for approval.

3.5.5 Each pump and its driver and controls Complies No Additional Clarification Drawing 8600-X-88446, Building shall be separated from the remaining fire Architectural Plan Fire Pumphouse Fire pumps and from the rest of the plant by Protection System, Rev. 3 / All rated fire barriers.

August 2011 Page A-50

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.5.15 Hydrants shall be installed approximately Complies with The exception to this requirement Drawing 8600-X-88448, Fire Loop and every 250 ft (76 m) apart on the yard Clarification is utilized at Callaway Plant by Laterals, Rev. 24 / All main system. A hose house equipped providing equipment on two mobile with hose and combination nozzle and units. Each mobile unit has CA2112, "Fire Brigade Equipment FPE other auxiliary equipment specified in equipment equivalent to that of Inventory and Condition Checklist," RAI NFPA 24, Standard for the Installation of three hose houses. dated 1/6/06 / All 06 Private Fire Service Mains and Their Appurtenances, shall be provided at Calculation KC-27, "NFPA Code intervals of not more than 1000 ft (305 m) Conformance Review," Rev. 0 /

along the yard main system. Appendix A, Section 24 Exception: Mobile means of providing NFPA 24, "Standard for Outside hose and associated equipment, such as Protection," 1973 Edition / All hose carts or trucks, shall be permitted in lieu of hose houses. Where provided, such mobile equipment shall be equivalent to the equipment supplied by three hose houses.

August 2011 Page A-56

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.5.16 The fire protection water supply system Submit for NRC NRC approval is being requested None shall be dedicated for fire protection use Approval in Attachment L for the use of the only. fire protection water supply system for purposes other than fire Exception No. 1: Fire protection water protection.

supply systems shall be permitted to be used to provide backup to nuclear safety systems, provided the fire protection water supply systems are designed and maintained to deliver the combined fire and nuclear safety flow demands for the duration specified by the applicable analysis.

Exception No. 2: Fire protection water storage can be provided by plant systems serving other functions, provided the storage has a dedicated capacity capable of providing the maximum fire protection demand for the specified duration as determined in this section.

3.6 Standpipe and N/A N/A N/A - General statement; No N/A Hose Stations technical requirements 3.6.1 For all power block buildings, Class III Complies with Standpipe and hose systems in Calculation KC-27, "NFPA Code LIC-standpipe and hose systems shall be Clarification power block buildings comply with Conformance Review," Rev. 0 / 06 installed in accordance with NFPA 14, NFPA 14-1976 Edition except as Appendix A, Section 14 Standard for the Installation of Standpipe, identified below.

Private Hydrant, and Hose Systems. NFPA 14, "Standard for the Installation of Standpipe and Hose Systems," 1976 Edition / All August 2011 Page A-57

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.6.1 For all power block buildings, Class III Complies by Previous Per Section 7-2.3 of NFPA Calculation KC-27, "NFPA Code standpipe and hose systems shall be NRC Approval 14-1976 Edition, the valves in the Conformance Review," Rev. 0 /

installed in accordance with NFPA 14, main connection to automatic Appendix A, Section 14 Standard for the Installation of Standpipe, sources of water supply shall be Private Hydrant, and Hose Systems. open at all times. There are motor- NFPA 14, "Standard for the Installation operated valves that isolate the of Standpipe and Hose Systems," 1976 containment standpipes, which Edition / Sections 7-2.3 and 7-2.4 must be opened manually from the control room to allow water into the Letter SLNRC 81-050 from Petrick containment standpipe risers. Per (SNUPPS) to Denton (NRC) dated Page 9.5B-225 of the attachment June 29, 1981 / Attachment, Pages to SLNRC 81-050, To protect the 9.5B-225 and 9.5E-1 chloride sensitive piping and equipment from fire protection NUREG-0830, "Safety Evaluation system leakage, the standpipes Report Related to the Operation of inside the reactor building are Callaway Plant, Unit No. 1," dated normally dry. Control room October 1981 / Section 9.5.1.6 operator action is required to LIC-charge the standpipes. The 05 probability of a fire occurrence is greater during refueling and maintenance operations.

Personnel will, therefore, be available during these operations to take the necessary action in the event of a fire.

Per Page 9.5E-1 of the attachment to SLNRC 81-050, "Wet standpipes for power block fire hoses are designed in accordance with the requirements for Class II service of NFPA No. 14-1976.

Hose racks are located so that no more than 100 feet separates adjacent hose racks. Access to permit functioning of the fire brigade is adequately discussed in Appendix 9.5B.

August 2011 Page A-58

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document "The standpipe system for the containment is supplied from the fire main loop through a safety-grade containment penetration.

The containment standpipes are normally dry and may be charged by operator action at the control room." LIC-06 Per Page 29 of NUREG-0830, "Manual hose stations are located throughout the plant to ensure that an effective hose stream can be directed to any safety-related area in the plant. The standpipes are consistent with the requirements of NFPA 14, "Standard for the Installation of Standpipe and Hose Systems." Standpipes are 4- and 2-1/2-in. diameter pipe for multiple and single hose station supplies, respectively, Based on this evaluation, the staff concludes that the sprinkler and standpipe systems are adequate, meet the guidelines of Appendix A, Sections C.3.a and C.3.d, and are, therefore, acceptable."

The standpipe and hose system, as approved in the referenced SER, is still in the same configuration as that which was approved. There have been no plant modifications or other changes that would invalidate the basis for approval.

LIC-05 August 2011 Page A-59

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.6.1 For all power block buildings, Class III Submit for NRC Hose stations protecting the ESW NUREG-0830, "Safety Evaluation standpipe and hose systems shall be Approval pump house are fed by the ESW Report Related to the Operation of installed in accordance with NFPA 14, system, not the fire protection Callaway Plant, Unit No. 1," dated Standard for the Installation of Standpipe, water system. The NRC approved October 1981 / Section 9.5.1.6 Private Hydrant, and Hose Systems. the standpipe and hose system in NUREG-0830 but the approval did not specifically include this configuration. This approval is being clarified in Attachment T.

3.6.2 A capability shall be provided to ensure Complies with Standpipe and hose stations Calculation M-KC-452, "Hose Station FPE an adequate water flow rate and nozzle Clarification comply with the requirements of Adequacy," Rev. 0 / All RAI pressure for all hose stations. This this section, except for those 07 capability includes the provision of hose protecting the ESW pump house Calculation KC-27, "NFPA Code station pressure reducers where as identified below. Conformance Review," Rev. 0 /

necessary for the safety of plant industrial Appendix A, Section 24, Code Section fire brigade members and off-site fire 4-4.2 department personnel.

A capability shall be provided to ensure Submit for NRC Hose stations protecting the ESW NUREG-0830, "Safety Evaluation an adequate water flow rate and nozzle Approval pump house are fed by the ESW Report Related to the Operation of pressure for all hose stations. This system, not the fire protection Callaway Plant, Unit No. 1," dated capability includes the provision of hose water system. The NRC approved October 1981 / Section 9.5.1.6 station pressure reducers where the standpipe and hose system in necessary for the safety of plant industrial NUREG-0830 but the approval did fire brigade members and off-site fire not specifically include this department personnel. configuration. This approval is being clarified in Attachment T.

August 2011 Page A-60

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.8.1 Fire Alarm. Alarm initiating devices shall be installed Complies by Previous Sections 2223 and 2231 of NFPA Letter SLNRC 84-0014 from Petrick LIC-in accordance with NFPA 72, National NRC Approval 72D-1975 Edition requires (SNUPPS) to Denton (NRC) dated 05 Fire Alarm Code. Alarm annunciation adequate secondary and remotely- February 1, 1984 / Enclosure 10 shall allow the proprietary alarm system located equipment power supplies.

to transmit fire-related alarms, Page 9-3 of NUREG-0830 NUREG-0830, "Safety Evaluation supervisory signals, and trouble signals Supplement 3 states, "The SER Report Related to the Operation of to the control room or other constantly states that the plant fire detection Callaway Plant, Unit No. 1," dated attended location from which required system is installed in accordance October 1981 / Section 9.5.1.6 notifications and response can be with NFPA 72D. During its site initiated. Personnel assigned to the visit, the staff noted that the back- Calculation KC-27, "NFPA Code LIC-proprietary alarm station shall be up power supply may not meet the Conformance Review," Rev. 0 / 05 permitted to have other duties. The recommendations of NFPA 72D. Appendix A, Section 72D following fire-related signals shall be The applicant was unable to show transmitted: compliance, and verbally agreed to NFPA 72D, "Standard for the prepare an analysis showing how Installation, Maintenance, and Use of the existing primary/back-up power Proprietary Protective Signaling supply circuitry compares to the Systems for Watchman, Fire Alarm and requirements of NFPA 72D. Supervisory Service," 1975 Edition /

Sections 1232, 2223, and 2231 "By letter dated February 1, 1984, the applicant provided the comparison. The applicant's comparison indicated that the primary and secondary power supplies comply with the provision of NFPA 72D. In the event of loss of power to the remote panels, loss of automatic activation of some pre-action sprinklers would occur. Because the pre-action systems are continuously supervised, any loss of power would be alarmed in the control room. The Plant Technical Specifications would then require the establishment of a continuous fire watch. Because of the fire watch and the fact that the sprinkler systems remain operable manually, the staff finds this to be August 2011 Page A-65

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.9.1(1) NFPA 13, Standard for the Installation of Complies, with Required See implementation item identified CAR 200902315, "NFPA 805 Sprinkler Systems Action below. Transition - Site Organizations Support Tracking CAR" FPE Modification MP 12-0009 RAI 14 IMPLEMENTATION ITEMS:

11-805-091 The missing ceiling tiles in the suspended ceiling in fire compartments C-5 and C-6 will be replaced in order to ensure proper operation of sprinkler system SKC34, which is credited in the Fire PRA, in accordance with NFPA 13-1976 Edition. Configuration control on the ceiling tiles will be ensured.

11-805-094 Modification MP 12-0009 will be completed to modify the quick-response sprinkler heads installed at an angle in cable chases to a FPE configuration that is in accordance with the requirements of NFPA 13-1976 Edition. RAI 14 3.9.1(2) NFPA 15, Standard for Water Spray N/A Automatic and manual water N/A Fixed Systems for Fire Protection based suppression systems credited to meet the requirements of Chapter 4 are identified in Table 4-3. There are no Chapter 4 credited NFPA 15 systems.

3.9.1(3) NFPA 750, Standard on Water Mist Fire N/A Water mist fire protection systems N/A Protection Systems are not used at Callaway Plant.

3.9.1(4) NFPA 16, Standard for the Installation of N/A Foam-water sprinkler and foam- N/A Foam-Water Sprinkler and Foam-Water water spray systems are not used Spray Systems at Callaway Plant.

August 2011 Page A-76

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.9.3 All alarms from fire suppression systems Complies M-22KC series drawings and FSAR SP, Section 9.5.1.2.2.1, Rev. OL- FPE shall annunciate in the control room or Drawing J-1073-00052 identify that 14f / Paragraph 3 RAI other suitable constantly attended all waterflow alarms annunciate on 09 location. panels that connect to KC008, System Description 10466-M-00KC, which is located in the control "Fire Protection System Description,"

room. Rev. 4 / Section 3.1.3 Drawing J-1073-00059, "KC008 and KC365 4120 Addressable Network Fire Alarm System Graphic Command Center Arrangement Details," Rev. 3 /

All Drawing M-22KC01, "P&ID, Fire Protection Turbine Building," Rev. 21 /

All Drawing M-22KC02, "P&ID, Fire Protection System Sheet 2," Rev. 21 /

All Drawing M-22KC03, "P&ID, Fire Protection System Sheet 3," Rev. 24 /

All Drawing M-22KC05, "P&ID, Fire Protection System Sheet 5," Rev. 11 /

All Drawing M-22KC08, "P&ID, Fire Protection Preaction Sprinkler System Sheet 8," Rev. 11 / All Drawing M-22KC09, "P&ID, Fire Protection System," Rev. 0 / All Drawing J-1073-00052, "KC324 4120 Addressable Network Fire Alarm Control Panel System Operation Matrix," Rev. 4 / All August 2011 Page A-78

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.10.1(3) NFPA 2001, Standard on Clean Agent N/A Clean agent fire extinguishing N/A Fire Extinguishing Systems systems are not used at Callaway Plant.

3.10.2 Operation of gaseous fire suppression Complies M-22KC series drawings identify Drawing M-22KC04, "Fire Protection FPE systems shall annunciate and alarm in that all system actuation alarms Halon System P&ID Sheet 4," Rev. 7 / RAI the control room or other constantly annunciate on panels that connect All 09 attended location identified. to KC008, which is located in the control room. Drawing M-22KC06, "Fire Protection Halon System P&ID Sheet 6," Rev. 3 Drawing M-22KC04, "Fire Protection Halon System P&ID Sheet 7," Rev. 7 /

All Drawing J-1073-00059, "KC008 and KC365 4120 Addressable Network Fire Alarm System Graphic Command Center Arrangement Details," Rev. 3 /

All 3.10.3 Ventilation system design shall take into Complies No Additional Clarification Calculation KC-27, "NFPA Code account prevention from over- Conformance Review," Rev. 0 /

pressurization during agent injection, Appendix A, Section 12A adequate sealing to prevent loss of agent, and confinement of radioactive Calculation KC-43, "NFPA 805 Code contaminants. Comparison," Rev. 0 / Attachment 4 August 2011 Page A-84

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

Table B NFPA 805 Ch. 3 Transition NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.10.8 Positive mechanical means shall be N/A Carbon dioxide extinguishing N/A provided to lock out total flooding carbon systems are not used in the Power dioxide systems during work in the Block.

protected space.

3.10.9 The possibility of secondary thermal Complies with As identified in the NFPA Code Calculation KC-27, "NFPA Code FPE shock (cooling) damage shall be Clarification Conformance Review of NFPA Conformance Review," Rev. 0 / RAI considered during the design of any 12A, a full system discharge test Appendix A, Section 12A 13 gaseous fire suppression system, but was performed for all Halon particularly with carbon dioxide. systems as part of the initial acceptance testing. No thermal impacts were noted as a result of these system discharges.

August 2011 Page A-86

Attachment B: Revisions to Transition Report Attachment B - NEI 04-02 Table B Nuclear Safety Capability Assessment -

Methodology Review

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment B - NEI 04-02 TABLE B Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection

• Trip and close control for Pressurizer Backup Group B breaker (PG2201)

NRC approval for the design of the Auxiliary Shutdown Panel, and for the overall Alternate Shutdown Strategy to meet the requirements of 10 CFR 50 Appendix R, Section III.G.3, was provided in NUREG-0830, SER Supplement No. 3, Docket No, STN 50-483, May 1984, and in NUREG-0830, SER Supplement No. 4, Docket No, STN 50-483, October 1984. Clarification regarding this approval is requested in Attachment T of the Callaway Plant NFPA 805 License Amendment Request, LDCN 11-0012, Transition Report.

Enabling of the Auxiliary Shutdown Panel involves the transfer of control from the Main Control Room to RP118B through an operator action to manually position three isolation transfer switches and five control switches which are located on RP118B. Following activation of the Auxiliary Shutdown Panel, the plant operator is provided with the capability to control and monitor secondary side decay heat removal capability utilizing the Auxiliary Feedwater System, the capability to control Reactor Coolant System (RCS) pressure, and the capability to monitor critical RCS process parameters which are necessary to verify that natural circulation has been established in the RCS and that it is being successfully maintained thereafter.

The Auxiliary Shutdown Panel has been transitioned to NFPA 805 as the Primary Control Station for meeting the NSPC in the event of a fire that requires evacuation of the Main Control Room.

Note: NUREG-0830 Supplement 3 identifies the following for the Main Control Room evacuation fire event: Some operations require cutting a control power SSA cable at the equipment to ensure that a fault in the control room does not prevent certain equipment operation. These operations have been superseded by RAI NFPA 805 plant modifications which provide for the capability to isolate and transfer control of the fire affected component to the local control station, with 03 redundant fusing. These NFPA 805 modifications are included in Attachment S of the LAR. There are no NFPA 805 Recovery Actions that require cutting of control power cable. The NFPA 805 Recovery Actions associated with the capability to isolate and transfer control of the fire affected component to the local control station, with redundant fusing, are identified and evaluated as VFDRs since they do not occur at the Primary Control Station, RP118B.

Reference Documents Calculation KC-26, "Nuclear Safety Capability Assessment," Revision 0 NUREG-0830, SER Supplement No. 3, Docket No, STN 50-483, May 1984 NUREG-0830, SER Supplement No. 4, Docket No, STN 50-483, October 1984 August 2011 Page B-12

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment B - NEI 04-02 TABLE B Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Note: The Instrument Air System has not been credited or analyzed in the Callaway Plant NSCA and NPO. The initial circuit analysis and cable selection, and SSA the subsequent deterministic fire area assessment for NFPA 805 NSCA and NPO components was performed utilizing the following criteria with respect to RAI considerations for the availability of instrument air. Instrument air system pressure IS assumed to exist if it can have an adverse consequence (i.e., air 02 pressure exists to keep an AOV in the undesired position absent operator action [from Main Control Room or credited Recovery Action] to ensure the pilot SOV is deenergized). Instrument air system pressure IS NOT assumed to exist if it can have a beneficial effect (i.e., air pressure exists to keep or place an AOV in the desired position).

Reference Documents Calculation KC-26, "Nuclear Safety Capability Assessment," Revision 0 August 2011 Page B-70

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment B - NEI 04-02 TABLE B Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis

• Spurious operation, when only resulting from positive to positive (+ to +) and negative to negative (- to -) external DC hot shorts in ungrounded DC circuits is only postulated in the circuit identification and analysis for high/low pressure interface valves and high consequence Fire PRA valves (as defined by the Fire PRA).

• No credit is taken for self-healing of electrical failures.

• Multiple AC and DC grounds are postulated in the circuit identification and analysis. Multiple grounds in ungrounded AC or DC systems can result in clearing of fuses, or tripping of breakers."

"* The circuit identification and analysis does not screen out cables on the basis of jacket material, insulation material, shielding, and/or the cable being routed in a dedicated conduit. However, the deterministic NSCA area-by-area analyses may discount spurious operation based on the fire affected cable being routed in a dedicated conduit, and therefore being protected from external sources of voltage."

Circuit identification and analysis for the Callaway Plant NSCA does not include limiting assumptions as described in RIS 2004-03.

Comments:

The circuit analysis and cable selection performed for Callaway is consistent with the guidelines, criteria, and assumptions of NEI 00-01 Revision 2. However, it has become apparent that NEI 00-01 may be unclear to some individuals with respect to the guidance, criteria, and assumptions as pertaining to inter-cable hot shorts (i.e., direct inter-cable hot shorts - source cable to target cable, and indirect inter-cable hot shorts - source cable to target cable through a ground plane). As a consequence, Callaway is providing the following clarification in the NFPA 805 LAR to describe the Callaway circuit analysis and cable selection treatment for inter-cable hot shorts, inclusive of direct and indirect inter-cable hot shorts:

The Callaway circuit analysis and cable selection process includes that a positive DC or a negative DC inter-cable hot short can occur on the same target cable LIC-so as to result in the spurious operation of a non-high/low pressure interface component.

17 The Callaway circuit analysis and cable selection process excludes that a positive DC and a negative DC inter-cable hot short can occur on the same target cable so as to result in the spurious operation of a non-high/low pressure interface component.

Inter-cable hot shorts are considered by Callaway to occur from direct source cable(s) to target cable interactions or from indirect source cable(s) to target cable interactions through a ground plane (i.e., the ground plane could be established through any fire affected plant equipment, conduits, and/or raceways).

No distinction is made by Callaway between direct and indirect inter-cable hot shorts. The mechanism for the externally applied voltage source (i.e., hot short) to contact the target cable is treated as a black box.

Based on this treatment, a non-high/low pressure interface component cannot spuriously operate due to a single inter-cable hot short (positive DC or negative DC) so long as there are also no adequate sources of DC voltage originating within the target cable that could result in spurious operation of the non-high/low pressure interface component due to a combination of intra-cable short circuits and a single inter-cable hot short.

August 2011 Page B-114

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment B - NEI 04-02 TABLE B Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis The Callaway treatment is consistent with NRC Generic Letter 86-10, Question and Answer 5.3.1, and NEI 00-01 Revision 2, Figure 3.5.2-5. The Callaway treatment is also consistent with the test results from NUREG/CR-7100, SAND2012-0323P, Direct Current Electrical Shorting in Response to Exposure Fire (DESIREE-Fire): Test Results, specific to inter-cable hot shorts (Section 6.5.3). Note the test configuration for the inter-cable hot shorts from NUREG/CR- LIC-7100, SAND2012-0323P, as depicted in Figure A-54, was set up intentionally to obtain inter-cable hot shorts for the study, and is not representative of field 17 typical installations which may further reduce the likelihood of inter-cable hot shorts.

Multiple grounds (in ungrounded circuits) are considered in the Callaway circuit analysis and cable selection process with respect to the potential for loss of required power for ungrounded circuits. This approach is consistent with NEI 00-01 Revision 2, Figure 3.5.2-3.

Reference Documents Calculation KC-26, "Nuclear Safety Capability Assessment," Revision 0 August 2011 Page B-115

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment B - NEI 04-02 TABLE B Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref NEI 00-01 Section 3 Guidance 3.5.2.3 Circuit Failures Due to a This section provides guidance for analyzing the effects of a hot short on circuits for required safe shutdown equipment.

Hot Short A hot short is defined as a fire induced insulation breakdown between conductors of the same cable, a different cable or some other external source resulting in an undesired impressed voltage on a specific conductor. The potential effect of the undesired impressed voltage would be to cause equipment to operate or fail to operate in an undesired manner.

Consider the following specific circuit failures related to hot shorts as part of the post-fire safe shutdown analysis:

- A hot short between an energized conductor and a de-energized conductor within the same cable may cause a spurious actuation of equipment. The spuriously actuated device (e.g., relay) may be interlocked with another circuit that causes the spurious actuation of other equipment. This type of hot short is called a conductor-to-conductor hot short or an internal hot short.

- A hot short between any external energized source such as an energized conductor from another cable (thermoplastic cables only) and a de-energized conductor may also cause a spurious actuation of equipment. This is called a cable-to-cable hot short or an external hot short. Cable-to-cable hot shorts between thermoset cables are not postulated to occur pending additional research.

Applicability Comments Applicable None Alignment Statement Aligns Alignment Basis Callaway Plant Calculation KC-26, Section 8.0, Circuit Identification and Analysis, identifies the overall process utilized to perform circuit identification and analysis for the NSCA components identified as being required to satisfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 805.

From Section 8.2 of KC-26:

"e. Postulate the effects of open circuits, short circuits, and/or grounds upon the desired position(s) / function(s) for the component at-power and/or non-power, as applicable"

"* Multiple simultaneous circuit failures are postulated in the circuit identification and analysis (affecting multiple cables, affecting multiple conductors within cables). No limit is prescribed to the number or type circuit failures that are postulated to occur except as modified by the following:

August 2011 Page B-122

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment B - NEI 04-02 TABLE B Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis

• Spurious operation, when resulting only from properly sequenced three-phase to three-phase external hot shorts is only postulated in the circuit identification and analysis for high/low pressure interface valves and high consequence Fire PRA valves (as defined by the Fire PRA).

• Spurious operation, when only resulting from positive to positive (+ to +) and negative to negative (- to -) external DC hot shorts in ungrounded DC circuits is only postulated in the circuit identification and analysis for high/low pressure interface valves and high consequence Fire PRA valves (as defined by the Fire PRA)."

"* The circuit identification and analysis does not screen out cables on the basis of jacket material, insulation material, shielding, and/or the cable being routed in a dedicated conduit. However, the deterministic NSCA area-by-area analyses may discount spurious operation based on the fire affected cable being routed in a dedicated conduit, and therefore being protected from external sources of voltage."

Comments:

The circuit analysis and cable selection performed for Callaway is consistent with the guidelines, criteria, and assumptions of NEI 00-01 Revision 2. However, it has become apparent that NEI 00-01 may be unclear to some individuals with respect to the guidance, criteria, and assumptions as pertaining to inter-cable hot shorts (i.e., direct inter-cable hot shorts - source cable to target cable, and indirect inter-cable hot shorts - source cable to target cable through a ground plane). As a consequence, Callaway is providing the following clarification in the NFPA 805 LAR to describe the Callaway circuit analysis and cable selection treatment for inter-cable hot shorts, inclusive of direct and indirect inter-cable hot shorts:

The Callaway circuit analysis and cable selection process includes that a positive DC or a negative DC inter-cable hot short can occur on the same target cable so as to result in the spurious operation of a non-high/low pressure interface component.

LIC-The Callaway circuit analysis and cable selection process excludes that a positive DC and a negative DC inter-cable hot short can occur on the same target 17 cable so as to result in the spurious operation of a non-high/low pressure interface component.

Inter-cable hot shorts are considered by Callaway to occur from direct source cable(s) to target cable interactions or from indirect source cable(s) to target cable interactions through a ground plane (i.e., the ground plane could be established through any fire affected plant equipment, conduits, and/or raceways).

No distinction is made by Callaway between direct and indirect inter-cable hot shorts. The mechanism for the externally applied voltage source (i.e., hot short) to contact the target cable is treated as a black box.

Based on this treatment, a non-high/low pressure interface component cannot spuriously operate due to a single inter-cable hot short (positive DC or negative DC) so long as there are also no adequate sources of DC voltage originating within the target cable that could result in spurious operation of the non-high/low pressure interface component due to a combination of intra-cable short circuits and a single inter-cable hot short.

The Callaway treatment is consistent with NRC Generic Letter 86-10, Question and Answer 5.3.1, and NEI 00-01 Revision 2, Figure 3.5.2-5. The Callaway treatment is also consistent with the test results from NUREG/CR-7100, SAND2012-0323P, Direct Current Electrical Shorting in Response to Exposure Fire (DESIREE-Fire): Test Results, specific to inter-cable hot shorts (Section 6.5.3). Note the test configuration for the inter-cable hot shorts from NUREG/CR-August 2011 Page B-123

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment B - NEI 04-02 TABLE B Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis 7100, SAND2012-0323P, as depicted in Figure A-54, was set up intentionally to obtain inter-cable hot shorts for the study, and is not representative of field typical installations which may further reduce the likelihood of inter-cable hot shorts. LIC-17 Multiple grounds (in ungrounded circuits) are considered in the Callaway circuit analysis and cable selection process with respect to the potential for loss of required power for ungrounded circuits. This approach is consistent with NEI 00-01 Revision 2, Figure 3.5.2-3.

Reference Documents Calculation KC-26, "Nuclear Safety Capability Assessment," Revision 0 August 2011 Page B-124

Attachment D: Revisions to Transition Report Attachment D - Non-Power Operational Modes Transition

SECURITY-RELATED INFORMATION - WITHHOLD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 Ameren Missouri Callaway Plant NFPA 805 Transition Report FAQ 07-0040 Implementing Guidance F.3 - Perform Fire Area Assessments (Identify pinch points)

Identify locations where:

x Fires may cause damage to the equipment (and cabling) credited above, or x KSFs are achieved solely by crediting recovery actions, e.g., alignment of gravity feed.

Fire modeling may be used to determine if postulated fires in a fire area are expected to damage equipment (and cabling) thereby eliminating a pinch point.

To implement this guidance perform the following tasks:

x Determine if a single fire in the fire area can cause loss of success paths for a KSF.

x Conservatively, assume the entire contents of a fire area are lost. Document the loss of success paths. Specifically identify those areas that cause loss of all success paths for a KSF.

x If fire modeling is used to limit the damage in a fire area, document that fire modeling is credited and ensure the basis for acceptability of that model (location, type, and quantity of combustible, etc.) is documented. These critical design inputs should be maintained during outage modes. Fire modeling treatment should include an assessment of safety margin to account for uncertainties/accuracy of the fire model used.

Review A deterministic fire separation analysis (i.e., assuming full area burn) was performed as documented in Callaway Plant Calculation KC-26, Nuclear Safety Capability Assessment, to identify pinch points (i.e., areas where redundant equipment and cables credited for a given KSF fail due to fire damage). There is a total of eighty-one (81) fire areas at the Callaway Plant; however, for the purposes of performing the computerized deterministic NFPA 805 NSCA in SAFE-PB, Fire Areas A-16 and C-1 were each subdivided into two (2) unique analysis area IDs; and Fire Area RB-1 was subdivided into five (5) unique analysis area IDs. Consequently the deterministic NFPA 805 NSCA included the analysis of eighty-seven (87) fire areas.

PRA x Fifty (50) fire areas were found to have an adequate number of KSF success paths to RAI 04-f survive the entire contents loss of the fire area.

PRA x Thirty-seven (37) fire areas were found to have pinch points resulting in the potential RAI 04-f loss of one or more KSFs success paths.

Fire modeling was not utilized to eliminate identification of pinch point fire areas as part of the implementation process for the step F.3 guidance from FAQ 07-0040.

Callaway Plant aligns with FAQ 07-0040 implementing guidance, F.3, Perform Fire Area Assessments (Identify pinch points).

August 2011 Page D-6 SECURITY-RELATED INFORMATION - WITHHOLD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390

Ameren Missouri Callaway Plant NFPA 805 Transition Report by fire area, the NPO risk reduction actions to be completed, and the evaluation and definition of the Callaway Plant POSs that are considered HREs.

Fire Protection Program procedure APA-ZZ-00700, Fire Protection Program, contains an overview of the NPO requirements, the commitments for implementation of the NPO risk reduction actions required by KC-26, Nuclear Safety Capability Assessment, and a road map to identify the site specific implementing procedures used to implement the NPO requirements (Implementation Item 11-805-058).

x APA-ZZ-00741, Control of Combustible Materials, contains controls to establish the outage roving fire watches that includes the required scope for the NPO risk reduction actions (Implementation Item 11-805-059).

x APA-ZZ-00742, Control of Ignition Sources, contains controls to establish fire watches for the hot work activities including all plant operating states within the NPO scope.

x APA-ZZ-00703, Fire Protection Operability Criteria and Surveillance Requirements, LIC-07 contains the compensatory actions to be implemented should a fire protection system required to be operable during HRE periods be found to be impaired. In these cases continuous fire watches will be implemented in the affected systems areas (Implementation Item 11-805-061).

x EDP-ZZ-04044, Fire Protection Reviews contains guidance to ensure that changes to the fire protection program are reviewed for impact to the NPO requirements and risk reduction actions (Implementation Item 11-805-062).

x APA-ZZ-00315, Configuration Risk Management, contains discussion on risk due to fire, NFPA 805 and the NPO requirements as part of risk management (Implementation Item 11-805-063).

Guidance is contained within the outage control procedures to ensure that upon entry into the NPO plant operating states the outage roving fire watches are established. No specific requirements are necessary for the hot work controls because they are in place in all plant operating states. Additional guidance and controls are in place to ensure the HRE risk reduction tools are implemented prior to entry into a plant HRE. Guidance is also in place to monitor the plant state (T-Boil Times) to determine when the HRE is exited.

x OTN-BB-00002, Reactor Coolant System Draining, contains a check list of the HRE risk reduction actions and requirements to implement the HRE risk reduction actions as a pre-requisite to conduct of the procedure which results in RCS reduced inventory states (Implementation Item 11-805-064).

x OTN-BB-00002, Addendum 7, Raising RCS Level to 6 Inches Below the RX Vessel Flange, contains a decision point to check RCS level and T-Boil to determine if the HRE has been exited, then an HRE Risk Reduction Actions exit check list is provided (Implementation Item 11-805-065).

x OTN-BB-00001, Reactor Coolant System, contains a requirement to maintain a continuous action to monitor T-Boil until the HRE can be exited then an HRE Risk Reduction Actions exit check list is provided (Implementation Item 11-805-066).

x OTG-ZZ-00007, Refueling Preparation, Performance and Recovery, contains a requirement to maintain a continuous action to monitor T-Boil until the HRE can be exited then an HRE Risk Reduction Actions exit check list is provided (Implementation Item 11-805-067).

August 2011 Page D-10

Attachment H: Revisions to Transition Report Attachment H - NFPA 805 Frequently Asked Question Summary Table

Ameren Missouri Callaway Plant NFPA 805 Transition Report Table H NEI 04-02 FAQs Utilized in LAR Submittal Closure No. Rev. Title FAQ Ref. Memo 08-0052 0 Transient Fire Growth Rate and ML081500500 ML092120501 Control Room Non-Suppression ML091590505 07-0054* 1 Demonstrating Compliance with ML103510379 ML110140183 Chapter 4 of NFPA 805 09-0056 2 Radioactive Release Transition ML102810600 ML102920405 08-0057 3 New Shutdown Strategy ML100330863 ML100960568 MP 10-0059 5 NFPA 805 Monitoring ML120410589 ML120750108 RAI 01

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

This table includes FAQs that have not been approved by the NRC but are utilized in this submittal based on industry concurrence with the guidance contained therein:

MP

RAI 01

Attachment J: Revisions to Transition Report Attachment J - Fire Modeling V&V

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment J - Table J V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion x The correlation is used in the NUREG-1805 fire model, for which V&V was documented in x NUREG-1805, Chapter 3, NUREG-1824.

2004 x The correlation is documented in an Flame Height Calculates the vertical x NUREG-1824, Volume 3, authoritative publication of the SFPE Handbook extension of the flame 2007 (Method of Heskestad) of Fire Protection Engineering.

region of a fire.

x SFPE Handbook, 4th x The correlation is applied within the validated Edition, Chapter 2-1, range reported in NUREG-1824 or has been LIC-10 Heskestad, 2008 justified as acceptable by qualitative analysis or quantitative sensitivity analysis.

x The correlation is used in the NUREG-1805 fire model, for which V&V was documented in x NUREG-1805, Chapter 9, NUREG-1824.

2004 x The correlation is documented in an x NUREG-1824, Volume 3, Calculates the vertical authoritative publication of the SFPE Handbook 2007 Plume Centerline separation distance, based of Fire Protection Engineering.

Temperature on temperature, to a target x SFPE Handbook, 4th x The correlation is applied within the validated in order to determine the Edition, Chapter 2-1, (Method of Heskestad) range reported in NUREG-1824 or has been vertical extent of the ZOI. Heskestad, 2008 LIC-10 justified as acceptable by qualitative analysis or x NUREG/CR-6850, quantitative sensitivity analysis.

Appendix H - Damage x NUREG/CR-6850 generic screening damage Criteria, 2005 criteria is used, which is considered conservative.

August 2011 Page J-2

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment J - Table J V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion x The correlation is used in the NUREG-1805 fire model, for which V&V was documented in x NUREG-1805, Chapter 5.

NUREG-1824.

2004 x The correlation is documented in an Calculates the horizontal x NUREG-1824, , Volume 4, authoritative publication of the SFPE Handbook separation distance, based 2007 of Fire Protection Engineering.

Radiant Heat Flux on heat flux, to a target in x SFPE Handbook, 4th order to determine the x The correlation is applied within the validated (Point Source Method) edition, Chapter 3-10, horizontal extent of the range reported in NUREG-1824 or has been LIC-10 Beyler, C., 2008 ZOI. justified as acceptable by qualitative analysis or x NUREG/CR-6850, quantitative sensitivity analysis.

Appendix H - Damage x NUREG/CR-6850 generic screening damage Criteria, 2005 criteria is used, which is considered conservative.

x The correlation is used in the FIVE-Rev1 fire model.

x x FIVE-Rev1, Referenced by The correlation is documented in an authoritative EPRI Report 1002981, publication of the SFPE Handbook of Fire Protection Calculates the horizontal 2002 Engineering.

radius, based on Plume Radius temperature, of the plume x SFPE Handbook, 4th x The Heskestad centerline plume correlation V&V at a given height. The Edition, Chapter 2-1, is documented in NUREG-1824.

(Method of Heskestad) correlation is derived of the Heskestad, G., 2008 LIC-10 Heskestad centerline x The Heskestad correlation is applied within the plume correlation. x NUREG/CR-6850, validated range reported in NUREG-1824 or has Appendix H - Damage been justified as acceptable by qualitative Criteria, 2005 analysis or quantitative sensitivity analysis.

x NUREG/CR-6850 generic screening damage criteria is used, which is considered conservative.

August 2011 Page J-3

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment J - Table J V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion x The correlation is used in the NUREG-1805 fire x NUREG-1805, Chapter 2, model, for which V&V was documented in 2004 NUREG-1824.

Calculates the hot gas x NUREG-1824, Volume 3, x The correlation is documented in an Hot Gas Layer layer temperature for a 2007 authoritative publication of the SFPE Handbook (Method of MQH) room with natural of Fire Protection Engineering.

ventilation. x SFPE Handbook, 4th Edition, Chapter 3-6, x The correlation is applied within the validated Walton W. and Thomas, P., range reported in NUREG-1824 or has been LIC-10 2008 justified as acceptable by qualitative analysis or quantitative sensitivity analysis.

x The correlation is used in the NUREG-1805 fire x NUREG-1805, Chapter 2, model, for which V&V was documented in 2004 NUREG-1824.

Calculates the hot gas x NUREG-1824, Volume 3, x The correlation is documented in an Hot Gas Layer layer temperature for a 2007 authoritative publication of the SFPE Handbook (Method of Beyler) closed compartment with of Fire Protection Engineering.

no ventilation. x SFPE Handbook, 4th Edition, Chapter 3-6, x The correlation is applied within the validated Walton W. and Thomas, P., range reported in NUREG-1824 or has been LIC-10 2008 justified as acceptable by qualitative analysis or quantitative sensitivity analysis.

August 2011 Page J-4

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment J - Table J V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion x The correlation is used in the NUREG-1805 fire x NUREG-1805, Chapter 2, model, for which V&V was documented in 2004 NUREG-1824.

Hot Gas Layer Calculates the hot gas x NUREG-1824, Volume 3, x The correlation is documented in an (Method of Foote, layer temperature for a 2007 authoritative publication of the SFPE Handbook Pagni, and Alvares room with forced of Fire Protection Engineering.

ventilation. x SFPE Handbook, 4th

[FPA]) Edition, Chapter 3-6, x The correlation is applied within the validated Walton W. and Thomas, P., range reported in NUREG-1824 or has been LIC-10 2008 justified as acceptable by qualitative analysis or quantitative sensitivity analysis.

x The correlation is used in the NUREG-1805 fire x NUREG-1805, Chapter 2, model, for which V&V was documented in 2004 NUREG-1824.

Hot Gas Layer Calculates the hot gas x NUREG-1824, Volume 3, x The correlation is documented in an layer temperature for a 2007 authoritative publication of the SFPE Handbook (Method of Deal and room with forced of Fire Protection Engineering.

Beyler) ventilation. x SFPE Handbook, 4th Edition, Chapter 3-6, x The correlation is applied within the validated Walton W. and Thomas, P., range reported in NUREG-1824 or has been LIC-10 2008 justified as acceptable by qualitative analysis or quantitative sensitivity analysis.

August 2011 Page J-5

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment J - Table J V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion x The correlation is used in the FIVE-Rev1 fire x FIVE-Rev1, Referenced by model, for which V&V was documented in EPRI Report 1002981, NUREG-1824.

2002 Calculates the horizontal x The correlation is documented in an separation distance, based x NUREG-1824, Volume 4, authoritative publication of the SFPE Handbook Ceiling Jet on temperature at the 2007 of Fire Protection Engineering.

Temperature ceiling of a room, to a x SFPE Handbook, 4th x The correlation is applied within the validated (Method of Alpert) target in order to determine Edition, Chapter 2-2, Alpert, range reported in NUREG-1824 or has been LIC-10 the horizontal extent of the R., 2008 justified as acceptable by qualitative analysis or ZOI. quantitative sensitivity analysis.

x NUREG/CR-6850, Appendix H - Damage x NUREG/CR-6850 generic screening damage Criteria, 2005 criteria is used, which is considered conservative.

x FDS Version 5 x V&V of the FDS is documented in NIST Special x NIST Special Publication Publication 1018-5.

Hot Gas Layer 1018-5, Volume 2: x The V&V of FDS specifically for Nuclear Power Calculations using Fire Used to calculate the hot Verification Plant applications has also been documented in Dynamics Simulator gas layer temperatures for x NIST Special Publication NUREG-1824.

(Version 5) various compartments, and the layer height. 1018-5, Volume 3: x The models are applied within their validated Validation range reported in NUREG-1824 or have been LIC-10 x NUREG-1824, Volume 7, justified as acceptable by qualitative analysis or 2007 quantitative sensitivity analysis.

August 2011 Page J-6

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment J - Table J V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion x FDS Version 5 x V&V of the FDS is documented in NIST Special x NIST Special Publication Publication 1018-5.

Sprinkler Actuation Used to estimate sprinkler 1018-5, Volume 2: x The V&V of FDS (for ceiling jet temperature)

Calculation using Fire actuation timing based on Verification specifically for Nuclear Power Plant applications Dynamics Simulator ceiling jet temperature, has also been documented in NUREG-1824.

(Version 5) x NIST Special Publication velocity, and thermal 1018-5, Volume 3:

response of sprinkler. x The models are applied within their validated Validation range reported in NUREG-1824 or have been LIC-10 x NUREG-1824, Volume 7, justified as acceptable by qualitative analysis or 2007 quantitative sensitivity analysis FM x The smoke detection correlation is used in the RAI 01-f x NUREG-1805, Chapter 11, NUREG-1805 fire model.

2004 x Alperts ceiling jet correlation V&V is x NUREG-1824, Volume 4, documented in NUREG-1824.

Alpert Ceiling Jet used to 2007 x The correlation is applied within the validated Smoke Detection determine temperature and th x SFPE Handbook, 4 range reported in NUREG-1824 or has been LIC-10 Actuation Correlation Heskestad and Edition, Chapter 4-1, Custer justified as acceptable by qualitative analysis or (Method of Heskestad Delichatsios temperature to R., Meacham B., and quantitative sensitivity analysis.

and Delichatsios) smoke density for smoke detection timing estimates. Schifiliti, R., 2008 x The temperature to smoke density correlation is th documented in an authoritative publication of the x SFPE Handbook, 4 Edition, Chapter 2-2, Alpert, SFPE Handbook of Fire Protection R., 2008 Engineering.

FM RAI 01-c August 2011 Page J-7

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment J - Table J V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion x The sprinkler actuation correlation is used in the NUREG-1805 fire model.

x The correlation is documented in an Used to estimate sprinkler x NUREG-1805, Chapter 10, authoritative publication of the NFPA Fire actuation timing based on 2004 Protection Handbook.

Sprinkler Activation the Alpert ceiling jet x NFPA Handbook, 19th LIC-10 Correlation temperature, velocity, and x Alperts ceiling jet correlation V&V is Edition, Chapter 3-9, thermal response of documented in NUREG-1824.

Budnick, E., Evans, D., and sprinkler. Nelson, H., 2003 x The correlation is applied within the validated range reported in NUREG-1824 or has been LIC-10 justified as acceptable by qualitative analysis or quantitative sensitivity analysis.

x V&V of the CFAST code is documented in the NIST Special Publication 1086.

x NIST Special Publication 1086, 2008 x The V&V of CFAST specifically for Nuclear Power Plant applications has also been Evaluates the time at which x CFAST Version 6 documented in NUREG-1824.

Control Room control room abandonment Abandonment x NUREG-1824, Volume 6, x The models are applied within their validated is necessary based on Calculation using 2007 range reported in NUREG-1824 or have been LIC-10 smoke obscuration and CFAST justified as acceptable by qualitative analysis or average HGL temperature. x NUREG/CR-6850, quantitative sensitivity analysis.

Appendix H - Damage Criteria, 2005 x NUREG/CR-6850 generic screening damage criteria is used, which is considered conservative.

August 2011 Page J-8

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment J - Table J V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion x V&V of the CFAST code is documented in the NIST Special Publication 1086.

x NIST Special Publication Determine the upper and 1086, 2008 x The V&V of CFAST specifically for Nuclear lower gas layer Power Plant applications has also been temperatures for various x CFAST Version 6 documented in NUREG-1824.

Temperature Sensitive compartments, and the x NUREG-1824, Volume 6, x The models are applied within their validated Equipment Hot Gas layer height, for use in 2007 range reported in NUREG-1824 or have been Layer Study LIC-10 assessing damage to justified as acceptable by qualitative analysis or temperature sensitive x NUREG/CR-6850, quantitative sensitivity analysis.

equipment. Appendix H - Damage Criteria, 2005 x NUREG/CR-6850 generic screening damage criteria is used, which is considered conservative.

x FDS Version 5 x V&V of the FDS is documented in the NIST x NIST Special Publication Special Publication 1018-5.

1018-5, Volume 2: x The V&V of FDS specifically for Nuclear Power Verification Plant applications has also been documented in Determine the radiant heat NUREG-1824.

x NIST Special Publication Temperature Sensitive flux ZOI at which 1018-5, Volume 3: x The models are applied within their validated Equipment Zone of temperature sensitive Validation range reported in NUREG-1824 or have been Influence Study equipment will reach LIC-10 damage thresholds. x NUREG-1824, Volume 7, justified as acceptable by qualitative analysis or 2007 quantitative sensitivity analysis.

x NUREG/CR-6850, x NUREG/CR-6850 generic screening damage Appendix H - Damage criteria is used, which is considered Criteria, 2005 conservative.

August 2011 Page J-9

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment J - Table J V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion x FDS Version 5 x V&V of the FDS is documented in NIST Special x NIST Special Publication Publication 1018-5.

1018-5, Volume 2: x The V&V of FDS specifically for Nuclear Power Verification Plant applications has also been documented in Determine the point at NUREG-1824.

x NIST Special Publication which hot gas layer and Plume/Hot Gas Layer 1018-5, Volume 3: x The models are applied within their validated plume interact and Interaction Study Validation range reported in NUREG-1824 or have been establish limits for plume LIC-10 temperature application. x NUREG-1824, Volume 7, justified as acceptable by qualitative analysis or 2007 quantitative sensitivity analysis.

x NUREG/CR-6850, x NUREG/CR-6850 generic screening damage Appendix H - Damage criteria is used, which is considered Criteria, 2005 conservative.

August 2011 Page J-10

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment J - Table J V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion x Zukoski, E.E., Properties of Fire Plumes, Combustion Fundamentals of Fire, Cox, G., Ed., Academic Press, London, 1995 x Sargent, W.S., Natural Convection Flows and Associated Heat Transfer Processes in Room Fires, Ph.D. thesis, California Institute of Technology, LIC-10 Pasadena, CA 1983 x Cetegen, B.M.,

Determines a heat release Entrainment and Flame x The correlation is applied within the validated rate adjustment factor for Geometry of Fire Plumes, range reported in the referenced studies or has Corner and Wall HRR fires that are proximate to a Ph.D. thesis, California been justified as acceptable by qualitative wall or corner. Institute of Technology, analysis or quantitative sensitivity analysis.

Pasadena, CA, 1982 x Williamson, R.B.

Revenaugh, A. and Mowrer, F.W., Ignition Sources in Room Fire Tests and Some Implications for Flame Spread Evaluation, International Association of Fire Safety Science, Proceedings of the Third International Symposium, New York, pp. 657-666, 1991 August 2011 Page J-11

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment J - Table J V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion x The correlation is recommended by NUREG/CR-6850.

x NUREG/CR-6850, Correlation for Heat Appendix R, 2005 x The correlation is documented in an Release Rates of Used to correlate bench- th authoritative publication of the SFPE Handbook x SFPE Handbook, 4 Cables scale data to heat release of Fire Protection Engineering.

Edition, Chapter 3-1, rates from cable tray fires.

(Method of Lee) Babrauskas, 2008 x The correlation is applied to configurations similar to those reported in NBISR 85-3195 or LIC-10 x NBISR 85-3195, July 1985 has been justified as acceptable by qualitative analysis.

x The correlation is recommended by NUREG/CR-7010 and follows guidance set forth in NUREG/CR-6850.

Correlation for Flame Used to predict the growth x NUREG/CR-7010, Section x The FLASH-CAT model is validated in Spread over Horizontal and spread of a fire within a 9, 2010 NUREG/CR-7010, Section 9.2.3, through Cable Trays (FLASH- vertical stack of horizontal x NUREG/CR-6850, experimentally measured HRRs compared with CAT) cable trays. Appendix R, 2005 the predictions of the FLASH-CAT model.

x The model is applied to configurations similar to those reported NUREG/CR-7010 or has been LIC-10 justified as acceptable by qualitative analysis.

August 2011 Page J-12

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment J - Table J V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion x FDS Version 5 x V&V of the FDS model is documented in NIST Special Publication 1018-5.

x NIST Special Publication 1018-5, Volume 2: x Roby, et al. validated the smoke detector Verification algorithm against a number of fire scenarios and geometries and concluded that the algorithm x NIST Special Publication and FDS model are accurately predicting the FM 1018-5, Volume 3: activation times of the smoke detectors. RAI 01-c Validation Section 2.2.3 x Cleary, et al. concluded that multi-room fire x R. Roby, et al. A Smoke simulation with the FDS software can yield Detector Algorithm for environmental conditions a detector or sensor Large Eddy Simulation may experience during an actual fire.

Used to predict detector Modeling, National Institute x The model is applied to configurations similar to Smoke Detector activation based on smoke of Standards and those reported in Roby, et al. or has been Activation using Fire production and velocity, as Technology, Gaithersburg, justified as acceptable by qualitative analysis or Dynamics Simulator well as detector geometry Maryland, July 2007. NIST quantitative sensitivity analysis.

(Version 5) and optical response. GCR 07-911 x T. Cleary, et al. Fire Detector Performance Predictions in a Simulated Multi-Room Configuration.

In Proceedings of the 12th International Conference on Automatic Fire Detection (AUBE 01). National Institute of Standards and Technology, Gaithersburg, Maryland, March 2001.

NIST SP 965. 12 August 2011 Page J-13

Ameren Missouri Callaway Plant NFPA 805 Transition Report Table J-1

References:

1. NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, U.S. Nuclear Regulatory Commission, Washington, DC, September 2005.

2. NUREG/CR-7010, Cable Heat Release, Ignition, and Spread in Tray Installations During Fire (CHRISTIFIRE), Volume 1: Horizontal Trays, Draft Report for Comment, United States Nuclear Regulatory Commission, October, 2010.

3. The SFPE Handbook of Fire Protection Engineering, 4th Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2008.

4. The NFPA Fire Protection Handbook, 19th Edition, A. E. Cote, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2003.

5. Peacock, R.D., Jones, W.W., Reneke, P.A., and Forney, G.P., CFAST - Consolidated Model of Fire Growth and Smoke Transport (Version

6) User's Guide, NIST Special Publication 1041, National Institute of Standards and Technology, Gaithersburg, MD, December 2005.

6. NIST Special Publication 1086, CFAST - Consolidated Model of Fire Growth and Smoke Transport (Version 6) Software Development and Model Evaluation Guide, National Institute of Standards and Technology, Gaithersburg, MD, December 2008.

7. NIST Special Publication 1018-5, Fire Dynamics Simulator (Version 5) Technical Reference Guide, Volume 2: Verification, National Institute of Standards and Technology, October 29, 2010

8. NIST Special Publication 1018-5, Fire Dynamics Simulator (Version 5) Technical Reference Guide, Volume 3: Validation, National Institute of Standards and Technology, October 29, 2010

9. Fire Modeling Guide for Nuclear Power Plant Applications, EPRI 1002981, FINAL REPORT, August 2002.

10. Inspection Manual Chapter (IMC) 0609, Appendix F, Fire Protection Significance Determination Process, Issue Date 02/28/05. LIC-10

11. R. Roby, et.al. A Smoke Detector Algorithm for Large Eddy Simulation Modeling, National Institute of Standards and Technology, Gaithersburg, Maryland, July 2007. NIST GCR 07-911.

FM

12. T. Cleary, et.al. Fire Detector Performance Predictions in a Simulated Multi-Room Configuration. In Proceedings of the 12th International RAI -01i Conference on Automatic Fire Detection (AUBE 01). National Institute of Standards and Technology, Gaithersburg, Maryland, March 2001.

NIST SP 965. 12.

13. Lee, B.T., NBISR 85-3195, Heat Release Rate Characteristics of Some Combustible Fuel Sources in Nuclear Power Plants, July 1985. LIC-10 August 2011 Page J-14

Attachment L: Revisions to Transition Report Attachment L - NFPA 805 Chapter 3 Requirements for Approval (10 CFR 50.48(c)(2)(vii))

Ameren Missouri Callaway Plant NFPA 805 Transition Report Approval Request 1 In accordance with 10 CFR 50.48(c)(2)(vii) Performance-based methods, the fire protection program elements and minimum design requirements of Chapter 3 may be subject to the performance-based methods permitted elsewhere in the standard.

In accordance with NFPA 805 Section 2.2.8, the performance-based approach to satisfy the nuclear safety, radiation release, life safety, and property damage/business interruption performance criteria requires engineering analyses to evaluate whether the performance criteria are satisfied.

In accordance with 10 CFR 50.48(c)(2)(vii), the engineering analysis performed shall determine that the performance-based approach utilized to evaluate a variance from the requirements of NFPA 805 Chapter 3:

(A) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; (B) Maintains safety margins; and (C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire nuclear safety capability).

Ameren Missouri requests formal approval of performance based exceptions requirements in Chapter 3 of NFPA 805 as follows:

NFPA 805, Section 3.5.16 NFPA 805, Section 3.5.16 states:

The fire protection (FP) water supply system shall be dedicated for fire protection use only.

Exception No. 1: Fire protection water supply systems shall be permitted to be used to provide backup to nuclear safety systems, provided the fire protection water supply systems are designed and maintained to deliver the combined fire and nuclear safety flow demands for the duration specified by the applicable analysis.

Exception No. 2: Fire protection water storage can be provided by plant systems serving other functions, provided the storage has a dedicated capacity capable of providing the maximum fire protection demand for the specified duration as determined in this section.

Contrary to the requirements of NFPA 805 Section 3.5.16, the Shift Manager/Control Room Supervisor (CRS) may approve use of fire protection system water for plant evolutions other than fire protection under the following conditions:

x Shift Manager/CRS approval is obtained and documented.

x A Fire Protection Impairment is generated to document the approvals, intended usage FPE and administrative controls in place using the fire protection impairment program (FPIP).

RAI 11

x Both fire water storage tanks are functional and have sufficient tank level margin based on the anticipated usage to remain functional during usage.

x Fire Water storage tank water level will be monitored to ensure the fire water storage tanks level remains above 260,000 gallons during use.

August 2011 Page L-2

Ameren Missouri Callaway Plant NFPA 805 Transition Report x Controls/communications are in place to ensure the non-fire protection system water demand can be secured immediately if a fire occurs.

x The non-fire protection system water demand must be less than 250 gpm.

Basis for Request:

The use of the fire protection water for these non-fire protection system water demands would have no adverse impact on the ability of the fire protection system to provide required flow and pressure, based on the following facts:

x The 250 gpm limitation is less than the hose stream postulated in determining fire suppression water flow requirements (a minimum of 500 gpm); therefore, there is no adverse impact on the flow and pressure available to any automatic water based suppression systems.

x Monitoring of fire water storage tank levels ensures the two tanks water volume will be FPE maintained above the procedurally required limit of 260,000 gallons. RAI 11 x Personnel utilizing the fire protection water will be in contact with the Control Room therefore ensuring the ability to secure the non-fire protection system water demand should a fire occur or tank level approach the procedurally required limit. Based on the FPE above controls adequate water flow will be available for the manual fire suppression RAI 11 demands when needed.

Nuclear Safety and Radiological Release Performance Criteria:

The use of fire protection water for non-FP plant evolutions is an occurrence requiring Shift Manager/CRS review and concurrence. The flow limitations ensure that there is no impact on the ability of the automatic suppression systems to perform their functions. The ability to isolate the non-fire protection flows ensures there is no impact on manual fire suppression efforts.

Therefore, there is no impact on the nuclear safety performance criteria.

The use of fire protection water for plant evolutions other than fire protection has no impact on the radiological release performance criteria. The radiological release performance criteria are satisfied based on the determination of limiting radioactive release (Attachment E), which is not affected by impacts on the fire protection system due to use of fire protection water for non-fire protection purposes.

Safety Margin and Defense-in-Depth:

The use of the fire water system, including the use of hydrants and hose, for non-fire protection uses does not impact fire protection defense-in-depth. The fire pumps have the excess capacity to supply the demands of the fire protection system in addition to the non-fire protection uses as identified above. This does not result in compromising automatic or manual fire suppression functions, fire suppression for systems and structures, or the nuclear safety capability assessment. Since both the automatic and manual fire suppression functions are maintained, defense-in-depth is maintained.

The methods, input parameters, and acceptance criteria used in this analysis were reviewed against those used for NFPA 805 Chapter 3 acceptance. The methods, input parameters, and acceptance criteria used to calculate flow requirements for the automatic and manual suppression systems were not altered. Therefore, the safety margin inherent in the analysis for the fire event has been preserved.

August 2011 Page L-3

Ameren Missouri Callaway Plant NFPA 805 Transition Report

==

Conclusion:==

NRC approval is requested for approval of the temporary use of the fire protection water supply with the following restrictions:

x Shift manager/CRS approval is obtained and documented; x A Fire Protection Impairment is generated to document the approvals, intended usage and administrative controls in place using the fire protection impairment program (FPIP).

FPE x Both fire water storage tanks are functional and have sufficient tank level margin based RAI 11 on the anticipated usage to remain functional during usage.

x Fire Water storage tank water level will be monitored to ensure the fire water storage tanks level remains above 260,000 gallons during use.

x Controls/communications are in place to ensure the non-fire protection water demand can be secured immediately if a fire occurs; x The non-fire protection system water demand must be less than 250 gpm.

The engineering analysis determined that the performance-based approach utilized to evaluate a variance from the requirements of NFPA 805 Chapter 3:

(A) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; (B) Maintains safety margins; and (C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire nuclear safety capability).

August 2011 Page L-4

Ameren Missouri Callaway Plant NFPA 805 Transition Report Approval Request 3 In accordance with 10 CFR 50.48(c)(2)(vii), Performance based methods, the fire protection program elements and minimum design requirements of Chapter 3 may be subject to the performance-based methods permitted elsewhere in the standard.

In accordance with NFPA 805 Section 2.2.8, the performance-based approach to satisfy the nuclear safety, radiation release, life safety, and property damage/business interruption performance criteria requires engineering analyses to evaluate whether the performance criteria are satisfied.

In accordance with 10 CFR 50.48(c)(2)(vii), the engineering analysis performed shall determine that the performance-based approach utilized to evaluate a variance from the requirements of NFPA 805 Chapter 3:

(A) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; (B) Maintains safety margins; and (C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire nuclear safety capability). LIC-15 Ameren Missouri requests formal approval of performance based exceptions to the requirements in Chapter 3 of NFPA 805 as follows:

NFPA 805, Section 3.2.3(1)

NFPA 805, Section 3.2.3(1) states:

Procedures shall be established for implementation of the fire protection program. In addition to procedures that could be required by other sections of the standard, the procedures to accomplish the following shall be established:

Inspection, testing, and maintenance for fire protection systems and features credited by the fire protection program.

Callaway Plant will utilize performance based methods to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features required by NFPA 805. Performance-based inspection, testing, and maintenance frequencies will be established as described in Electric Power Research Institute (EPRI) Technical Report TR-1006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features", Final Report, July 2003.

Basis for Request:

NFPA 805 Section 2.6, Monitoring, requires that A monitoring program shall be established to ensure that the availability and reliability of the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria. Monitoring shall ensure that the assumptions in the engineering analysis remain valid.

NFPA 805 Section 2.6.1, Availability, Reliability, and Performance Levels, requires that Acceptable levels of availability, reliability, and performance shall be established.

NFPA 805 Section 2.6.2 requires that Methods to monitor availability, reliability, and performance shall be established. The methods shall consider the plant operating experience and industry operating experience.

August 2011 Page L-8

Ameren Missouri Callaway Plant NFPA 805 Transition Report The scope and frequency of the inspection, testing, and maintenance activities for fire protection systems and features required in the fire protection program have been established based on the previously approved Technical Specifications / License Controlled Documents and appropriate NFPA codes. This request does not involve the use of the EPRI Technical Report TR-1006756 to establish the scope of those activities as that is determined by the required systems review identified in Table 4-3.

This request is specific to the use of EPRI Technical Report TR-1006756 to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features credited by the fire protection program. As stated in EPRI Technical Report TR-1006756 Section 10.1, The goal of a performance-based surveillance program is to adjust test and inspection frequencies commensurate with equipment performance and desired reliability.

This goal is consistent with the stated requirements of NFPA 805 Section 2.6. The EPRI Technical Report TR-1006756 provides an accepted method to establish appropriate inspection, LIC-15 testing, and maintenance frequencies which ensure the required NFPA 805 availability, reliability, and performance goals are maintained.

The target tests, inspections and maintenance will be those activities for the NFPA 805 required Fire Protection systems and features. The reliability and frequency goals will be established to ensure the assumptions in the NFPA 805 engineering analysis remain valid. The failure criterion will be established based on the required Fire Protection systems and features credited functions and will ensure those functions are maintained. Data collection and analysis will follow the Technical Report TR-1006756 document guidance. The failure probability will be determined based on the Technical Report TR-1006756 guidance and a 95% confidence level will be utilized. The performance monitoring will be performed in conjunction with the Monitoring program required by NFPA 805 section 2.6 and it will ensure site specific operating experience is considered in the monitoring process. The following is a flow chart that identifies the basic process that will be utilized.

August 2011 Page L-9

Ameren Missouri Callaway Plant NFPA 805 Transition Report LIC-15 EPRI TR-1006756 - Figure 10-1 Flowchart for Performance-Based Surveillance Program August 2011 Page L-10

Ameren Missouri Callaway Plant NFPA 805 Transition Report Acceptance Criteria Evaluation:

Nuclear Safety and Radiological Release Performance Criteria:

Use of performance based test frequencies established per EPRI Technical Report TR-1006756 methods combined with NFPA 805 Section 2.6, Monitoring Program, will ensure that the availability and reliability of the fire protection systems and features are maintained to the levels assumed in the NFPA 805 engineering analysis. Therefore, there is no adverse impact to Nuclear Safety Performance Criteria by the use of the performance based methods in EPRI Technical Report TR-1006756.

The radiological release performance criteria are satisfied based on the determination of limiting radioactive release (Refer to Attachment E of this LAR). FP Systems and features are credited as part of that evaluation. Use of performance based test frequencies established per EPRI Technical Report TR-1006756 methods combined with NFPA 805 Section 2.6, Monitoring Program will ensure that the availability and reliability of the fire protection systems and features are maintained to the levels assumed in the NFPA 805 engineering analysis which includes those assumptions credited to meet the Radioactive Release performance criteria. Therefore, there is no adverse impact to Radioactive Release performance criteria.

Safety Margin and Defense-in-Depth: LIC-15 Use of performance based test frequencies established per EPRI Technical Report TR-1006756 methods combined with NFPA 805 Section 2.6, Monitoring Program will ensure that the availability and reliability of the fire protection systems and features are maintained to the levels assumed in the NFPA 805 engineering analysis which includes those assumptions credited in the Risk Evaluation safety margin discussions. In addition, the use of these methods in no way invalidates the inherent safety margins contained in the codes used for design and maintenance of fire protection systems and features. Therefore, the safety margin inherent and credited in the analysis has been preserved.

The three echelons of defense-in-depth described in NFPA 805 section 1.2 are 1) to prevent fires from starting (combustible/hot work controls), 2) rapidly detect, control and extinguish fires that do occur thereby limiting damage (fire detection systems, automatic fire suppression, manual fire suppression, pre-fire plans), and 3) provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed (fire barriers, fire rated cable, success path remains free of fire damage, recovery actions).

Echelon 1 is not affected by the use of EPRI Technical Report TR-1006756 methods. Use of performance based test frequencies established per EPRI Technical Report TR-1006756 methods combined with NFPA 805 Section 2.6 Monitoring Program, will ensure that the availability and reliability of the fire protection systems and features credited for DID are maintained to the levels assumed in the NFPA 805 engineering analysis. Therefore, there is no adverse impact to echelons 2 and 3 for the defense in depth.

August 2011 Page L-11

Ameren Missouri Callaway Plant NFPA 805 Transition Report

==

Conclusion:==

NRC approval is requested for use of the performance based methods contained in Electric Power Research Institute (EPRI) Technical Report TR-1006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features", Final Report, July 2003 to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features required by NFPA 805. As described above, this approach is considered acceptable because it:

(A) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; (B) Maintains safety margins; and (C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire nuclear safety capability).

August 2011 Page L-12

Attachment S: Revisions to Transition Report Attachment S - Plant Modifications and Items to be completed during Implementation

Ameren Missouri Callaway Plant NFPA 805 Transition Report Tables S-1 and S-2, Plant Modifications Completed and Committed, respectively. Each of these tables provided below includes a description of the modifications along with the following information:

x A problem statement, x Risk ranking of the modification based on estimated impact on the Fire PRA results (see legend),

x An indication if the modification is currently included in the FPRA, x Compensatory Measure in place, and x A risk-informed characterization of the modification and compensatory measure.

x The following legend should be used when reviewing the tables:

o High = Modification would have an appreciable impact on reducing overall fire CDF.

o Medium = Modification would have a measurable impact on reducing overall fire CDF.

o Low = Modification would have either an insignificant or no impact on reducing overall fire CDF.

Table S Plant Modifications Completed In Comp Item Rank Problem Statement Proposed Modification Risk Informed Characterization FPRA Measure 07-0066 M Buried carbon steel ESW system The buried carbon steel ESW Y No Cables affect ESW cooling from piping needed replacement. As piping was replaced with high the UHS cooling towers, part of this piping modification, density polyethylene (HDPE) potentially failing both trains, but relocate cables currently in piping. During the piping could be mitigated by a recovery nonconformance with 20 foot replacement the cabling action (and the recovery action is separation criteria. associated with EFTE0067A no longer needed). This is judged and 68A was relocated to to have a medium impact on risk.

restore the required 20 foot separation criteria. Compensatory measure: None; modification complete August 2011 Page S-2

Ameren Missouri Callaway Plant NFPA 805 Transition Report Table S Plant Modifications Completed In Comp Item Rank Problem Statement Proposed Modification Risk Informed Characterization FPRA Measure 10-0032 H Risk metrics indicated that Installed a non-safety related Y4 No Fire PRA credits this modification additional defense in depth was AFW pump as diverse AFW for decay heat removal warranted for the AFW system. backup supply to the safety redundancy.

This modification provides margin related motor driven and for AFW system MSPI metrics. turbine driven pumps. Compensatory measure: None; modification complete 4

10-0038 H Improve Callaway Plants Install four non-safety related Y No Fire PRA credits this modification defense in depth to mitigate the diesel generators (8 MW) at the for electrical power redundancy consequences from a potential electric cooperative substation.

Station Black Out (SBO). Provide Either the electric cooperative Compensatory measure: None; an alternate emergency source of substation or the 4 non-safety modification installed. Additional power that is diverse from the diesel generators will be able to changes forthcoming that do not Emergency Diesel Generators power either Safety Related affect FPRA.

and offsite sources. bus in the event of a loss of AC power and failure of the Emergency Diesel Generators.

FPE RAI 14 Table S-2, Items provided below are those modifications that will be completed prior to the implementation of the new NFPA 805 FP program.

Currently open modifications will be field completed no later than June 30, 2013. Appropriate compensatory measures for any incomplete NFPA 805 related modifications will be maintained until the modifications are complete.

4 Refer to associated implementation item in Table S-3.

August 2011 Page S-3

Ameren Missouri Callaway Plant NFPA 805 Transition Report Table S Plant Modifications Committed In Comp Item Rank Problem Statement Proposed Modification Risk Informed Characterization FPRA Measure 05-3029 Install lower amperage fuses to Install lower amperage fuses for N No This modification ensures there are prevent damage to 14 AWG various 14 AWG control circuits in no secondary fires. NUREG/CR-cables and prevent secondary the MCR. The majority of the 6850 methodology does not fires from occurring in the MCR. modification centers around the address secondary fires, but the trip circuit fuses on NB, NG, PA, issue of secondary fires was raised PB, and PG system breakers. during the pilot plant RAI process.

Secondary fires are not modeled in the Callaway Fire PRA and an assessment of risk was not performed.

Compensatory measure for NFPA 805: In accordance with station procedures, appropriate compensatory measures will be established when the NFPA 805 fire protection program becomes effective and remain in effect until this modification is complete.

Compensatory measure for Current Fire Protection Licensing Basis:

None; the MCR is deterministically compliant with the Current Fire Protection Licensing Basis.

August 2011 Page S-4

Ameren Missouri Callaway Plant NFPA 805 Transition Report Table S Plant Modifications Committed In Comp Item Rank Problem Statement Proposed Modification Risk Informed Characterization FPRA Measure 07-0151 L During a fire in the main control Install redundant fuses and Y Y PRA assumes that after a fire in the room (MCR), selected cables to isolation switches for MCR FPIP 14050 main control room, the B train B train related equipment fed evacuation procedure OTO-ZZ- components are operable from the from NB02 will be isolated to 00001. auxiliary shutdown panel without prevent a multi-spurious hot requiring replacement of fuses.

short from stopping or starting Compensatory measure for NFPA safety equipment. Circuits that 805: In accordance with station have isolation switches but lack procedures, appropriate redundant fuses are included in compensatory measures will be this modification. This established when the NFPA 805 fire modification will eliminate credit protection program becomes previously taken to have effective and remain in effect until operators replace potentially this modification is complete.

blown fuses prior to the NFPA Compensatory measure for Current 805 transition.

Fire Protection Licensing Basis:

None; the MCR is deterministically compliant with the Current Fire Protection Licensing Basis.

August 2011 Page S-5

Ameren Missouri Callaway Plant NFPA 805 Transition Report Table S Plant Modifications Committed In Comp Item Rank Problem Statement Proposed Modification Risk Informed Characterization FPRA Measure 09-0025 M A fire in fire areas A-1, A-2, A- To protect against multiple Y Y This is judged to be a moderate risk 4, A-8, A-16 (analysis area A- spurious scenarios, the solution is FPIP 14050 improvement. In A-1, A-2, A-4, A-8, 16S), A-27, C-18, C-21, C-22, to run a single wire in a protected A-16 (analysis area A-16S), A-27, C-23, C-24, C-30, C-33, and metal jacket such that spurious C-18, C-21, C-22, C-23, C-24, C-30, RB-1 (analysis areas RB1, valve opening due to a hot short C-33, and RB-1 (analysis areas RB2, RB3, and RB4) could affecting the valve control circuit RB1, RB2, RB3, and RB4),

cause EJHV8811A and/or B to is eliminated for these fire areas. EJHV8811A/B are assumed to not spuriously open due to direct have potential for spurious opening valve control cable damage and due to valve control cable damage begin draining the RWST to the because of the modification.

containment emergency EJHV8811A/B can still spuriously sumps. open if the MCC which powers the valve is involved in the fire, or in response to a valid or spurious SI signal concurrent with a spurious RWST Low level signal.

Compensatory measure for NFPA 805: In accordance with station procedures, appropriate compensatory measures will be established when the NFPA 805 fire protection program becomes effective and remain in effect until this modification is complete.

Compensatory measure for Current Fire Protection Licensing Basis :

None; fire areas A-1, A-2, A-4, A-8, A-16 (analysis area A-16S), A-27, C-18, C-21, C-22, C-23, C-24, C-30, C-33, and RB-1 (analysis areas RB1, RB2, RB3, and RB4) are deterministically compliant with the Current Fire Protection Licensing Basis.

August 2011 Page S-6

SECURITY-RELATED INFORMATION - WITHHOLD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 Ameren Missouri Callaway Plant NFPA 805 Transition Report Table S Plant Modifications Committed In Comp Item Rank Problem Statement Proposed Modification Risk Informed Characterization FPRA Measure 12-0009 L Quick response sprinkler heads Quick response sprinkler heads Y Y The risk from this condition is low.

are installed in cable chases A- in cable chases A-11, C-30, and FPIP 21315 While the sprinkler heads do not 11, C-30, and C-31. Due to the C-31 will be modified to be in explicitly meet NFPA code they are piping configuration, the quick accordance with the applicable installed and functional and will response sprinkler heads were requirements of NFPA 13-1976 activate in the event of a fire and installed at an angle relative to edition. provide full coverage within the fire the ceiling, as opposed to being area.

parallel to it; the latter of which Compensatory measure for NFPA is typical. 805: As required by the approved Fire Protection Program an hourly FPE roving fire watch has been RAI established for fire areas A-11, C-14 30, and C-31 which will remain in place until the sprinkler system is modified to be compliant with the NFPA code.

Compensatory measure for Current Fire Protection Licensing Basis: As required by the approved Fire Protection Program an hourly roving fire watch has been established for fire areas A-11, C-30, and C-31 which will remain in place until the sprinkler system is modified to be compliant with the NFPA code.

August 2011 Page S-7 SECURITY-RELATED INFORMATION - WITHHOLD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390

Ameren Missouri Callaway Plant NFPA 805 Transition Report Table S-3 Implementation Items Item Unit Description LAR Section / Source 11-805-049 1 Section 4.1.5.b of APA-ZZ-00741 will be revised to address that cribbing timbers 6 in. by 6 in. or 4.1.2 and Attachment A larger are not required to be fire-retardant treated.

11-805-050 1 Drawing E-2R8900 and procedure EDP-ZZ-04044 will be revised to require that, where wiring 4.1.2 and Attachment A must be installed above a suspended ceiling, it shall be of a type approved in FAQ 06-0022.

11-805-051 1 Section 4.1.3(c) of procedure APA-ZZ-00743, "Fire Team Organization and Duties," will be 4.1.2 and Attachment A revised to include the requirement that industrial fire brigade members shall have no other assigned normal plant duties that would prevent immediate response to a fire or other emergency as required.

11-805-052 1 Procedure APA-ZZ-00700 will be revised to identify that plant personnel who respond with the 4.1.2 and Attachment A industrial fire brigade are trained as to their responsibilities, potential hazards to be encountered, and interfacing with the industrial fire brigade.

11-805-053 1 OTO-ZZ-00001 and OTO-KC-00001 will be revised to incorporate credited Recovery Actions 4.2.1.3 and Attachment G consistent with Attachment C (Fire Area Transition).

11-805-055 1 Non-Power Operations risk management strategies from the NFPA 805 NSCA (Callaway Plant 4.2.1 and Attachment D Calculation KC-26, "Nuclear Safety Capability Assessment") and the FSAs for fire areas with identified KSF pinch points will be incorporated into the plant fire response procedure(s), plant outage management procedures, and plant operating procedure(s).

11-805-056 1 Confirmation that plant modification MP 07-0151 has adequately modified the control circuitry 4.2.4 and Attachment C for Emergency Diesel Generator NE02, such that local isolation/transfer/control capability for the Main Control Room fire evacuation scenario is maintained without having to replace fuses, cut wires, or perform other repair activities with consideration given to fire induced multiple simultaneous hot shorts, open circuits, and shorts to ground per the criteria of NEI 00-01, will be made. Confirmation that the modification is correctly implemented into procedure OTO-ZZ-00001 will be made.

11-805-058 1 APA-ZZ-00700, Fire Protection Program, will be revised to add NPO overview, definitions; 4.3.2 and Attachment D road map; and risk reduction requirements for all NPO, then HRE.

11-805-059 1 APA-ZZ-00741, Control of Combustible Materials, will be revised to add a section which 4.3.2 and Attachment D addresses outage roving fire watches with specific NPO scope.

11-805-061 1 APA-ZZ-00703, "Fire Protection Operability Criteria and Surveillance Requirements, contains 4.3.2 and Attachment D LIC-the compensatory actions to be implemented should a fire protection system required to be 07 operable during HRE periods be found to be impaired. In these cases continuous fire watches will be implemented in the affected systems areas.

August 2011 Page S-11

Ameren Missouri Callaway Plant NFPA 805 Transition Report Table S-3 Implementation Items Item Unit Description LAR Section / Source 11-805-073 1 The audit scope requirements contained in OQAM Section 18.8.e will be relocated to FSAR SP 4.6.2 Section 9.5.1 and revised to add the Monitoring Program assessment criteria identified in LIC-Section 4.6.2. Additionally the OQAM Section 18 will be revised to change the FP QA Audit 19 frequency from 2 years to 3 years.

11-805-074 1 Procedure ODP-ZZ-00002, "Equipment Status Control," Attachment 3, "Operability Evaluations," 4.2.1.2 will be revised to ensure the assumed nitrogen inventory as described in Section 4.2.1.2 Safe and Stable Conditions for the Plant, is maintained in the ASD N2 accumulator tanks.

11-805-075 1 In accordance with ODP-ZZ-0016E, "Operations Technicians Watchstation Practices and 4.2.1.2 Rounds," a form will be initiated to change the data points for Operations AutoTour to ensure the assumed nitrogen inventory as described in Section 4.2.1.2 Safe and Stable Conditions for the Plant, is maintained in the ASD N2 accumulator tanks.

11-805-076 1 The Fire Pre-Plan Manual will be revised as follows: 4.1.2 and Attachment A

• The fire pre-plan attachments will be revised where the radiation release criteria are applicable for gaseous and liquid effluent as described in Table E-1/E-2 to include effluent controls and monitoring.

• New Pre-Fire Plans will be added for C-36 and C-37.

• Two new Attachments will be added, for Temporary Structures Inside the PA and for Temporary Structures Outside the PA, and existing Fire Attack Guidelines will be combined into each attachment.

11-805-077 1 FPP-ZZ-00009, "Fire Protection Training Program," will be revised to include the containment 4.4.2 and Attachment E and monitoring of fire suppression agents and products of combustion in potentially contaminated areas.

11-805-078 1 FPP-ZZ-00009, "Initial Traning Course Agenda," will be revised to include the containment and 4.4.2 and Attachment E monitoring of fire suppression agents and products of combustion in potentially contaminated areas.

11-805-079 1 FPP-ZZ-00009, "Retraining Courses and Activities," will be revised to include the containment 4.4.2 and Attachment E and monitoring of fire suppression agents and products of combustion in potentially contaminated areas.

11-805-080 1 Section 6 of HTP-ZZ-05006, Fire Involving Radioactive Material or Entry into the Radiological 4.4.2 and Attachment E Controlled Area, will be revised to address Radiation Protection actions for monitoring and control of potentially contaminated effluents.

August 2011 Page S-13

Ameren Missouri Callaway Plant NFPA 805 Transition Report Table S-3 Implementation Items Item Unit Description LAR Section / Source 11-805-088 1 Configuration control mechanisms for the Fire PRA and NSCA will be revised to ensure the 4.2.1.4 and Attachment F basis for MSO inclusion/exclusion is maintained consistent with the current plant. The rationale for excluding generically identified MSOs from the Callaway Plant Fire PRA and Callaway Plant NSCA was documented in Callaway Plant Calculation 17671-002b, "Callaway NFPA 805 Fire PRA - MSO Expert Panel Report," and Callaway Plant Calculation KC-26, "Nuclear Safety Capability Assessment," respectively. Configuration control mechanisms will be reviewed to provide reasonable confidence that the exclusion basis remains valid.

11-805-089 1 The Monitoring program described in procedure EDP-ZZ-01101, Fire Protection Monitoring 4.6.2 MP Program Procedure, will be implemented after the safety evaluation issuance as part of the fire RAI protection program transition to NFPA 805. Ameren Missouri will implement a monitoring 01-a program in accordance with FAQ 10-0059 Rev. 5 during implementation.

11-805-090 1 In order to adequately reflect the calculated reliability of recovery actions in the Fire HRA, the 4.2.1.3 and Attachment G Fire HRA will be updated once procedure updates, plant modifications, and recovery action training are complete.

11-805-091 1 The missing ceiling tiles in the suspended ceiling in fire compartments C-5 and C-6 will be 4.1.2 and Attachment A replaced in order to ensure proper operation of sprinkler system SKC34, which is credited in the Fire PRA, in accordance with NFPA 13-1976 Edition. Configuration control on the ceiling tiles will be ensured.

11-805-092 1 An administrative control will be implemented to ensure that the breaker (PB0406) for DPAE02 4.2.1.4 and Attachment F is disabled open during at-power plant operation to address a potential MSO that may result in overfill of steam generators/overcooling of the RCS.

11-805-093 1 The current Fire PRA does include the phase #1 version of MP 10-0038, and the AEPS, MP 10- Attachment S 0032. As the modification packages are completed, the FPRA will be udpated if necessary to reflect the final configuration.

12-805-001 1 All of the items labeled with footnote 1 in PRA RAI 1, Table 1, will be completed: Attachment U PRA F&Os: 1-7, 1-14, 1-25, 1-3,1-13 and 1-15 RAI 01 These commitments involve updating the fire PRA to be consistent with upgrade items that were and implemented in the internal events PRA update (Revision 5) after the Callaway Plant NFPA-805 PRA LAR was submitted.

RAI 02 In addition, a self-assessment of the internal events PRA against the RG 1.200, Rev 2 clarifications and qualifications to determine if any gaps exist is in progress and will be completed, with any resolutions completed before transition to NFPA 805 occurs.

August 2011 Page S-15

Ameren Missouri Callaway Plant NFPA 805 Transition Report Table S-3 Implementation Items Item Unit Description LAR Section / Source

  7HFKQLFDOGHWDLOVGHPRQVWUDWLQJWKH&RQWURO5RRP$EDQGRQPHQWPRGHOVDUHXVHGZLWKLQWKH $WWDFKPHQW- FM YDOLGDWHGUDQJHRI185(*³9HULILFDWLRQDQG9DOLGDWLRQRI6HOHFWHG)LUH0RGHOVIRU1XFOHDU RAI 3RZHU3ODQW$SSOLFDWLRQV')LQDO5HSRUW$SULODVZHOODVDQ\MXVWLILFDWLRQRIPRGHOV 01-f RXWVLGHWKHUDQJHZLOOEHGRFXPHQWHGLQWKHQH[WXSGDWHWR&DOODZD\&DOFXODWLRQE

³&DOODZD\1)3$)LUH35$0DLQ&RQWURO5RRP)LUH$QDO\VLV

PRA

  7KHILUH35$/(5)PRGHOZLOOEHXSGDWHGWREHFRQVLVWHQWZLWKWKH,(35$/(5)PRGHO DQG$WWDFKPHQW:

RAI SKHQRPHQRORJ\VSOLWIUDFWLRQVDQGSUREDELOLWLHVGXULQJWKHQH[W)LUH35$XSGDWH

03 August 2011 Page S-16

Attachment T: Revisions to Transition Report Attachment T - Clarification of Prior NRC Approvals

Ameren Missouri Callaway Plant NFPA 805 Transition Report x Open and close control for Steam Generator D (4) Atmospheric Steam Dump Valve (ABPV0004) x Open and close control for Steam Generator D (4) AFW flow control valve from MDAFW Pump B (ALHV0005) x Open and close control for Essential Service Water to suction of TDAFP (ALHV0033) x TDAFP suction pressure indication (ALPI0026B) x Open and close control for TDAFP Governor Control valve (FCFV0313) x Open and close control for TDAFP Trip and Throttle valve (FCHV0312) x Pressurizer level indication (BBLI0460B) x Reactor Coolant System pressure indication (BBPI0406X) x Reactor Coolant System Loop 2 cold leg temperature indication (BBTI0423X) x Reactor Coolant System Loop 4 hot leg temperature indication (BBTI0443A) x Intermediate and source range neutron monitoring indication (SENI0061X and SENI0061Y) x Trip and close control for Pressurizer Backup Group B breaker (PG2201)

Request As part of this LAR submittal and transition to NFPA 805, it is requested that the NRC formally document as prior approval the physical design and capabilities for Auxiliary Shutdown Panel RP118B, including the specific components and features cited above.

The phased procedural approach that is discussed in the SER Supplement 4 approval has been revised as part of the NFPA 805 transition. Ameren Missouri seeks only to maintain the approval of the original design of the ASP and its physical capabilities. The NSCA has been performed under the transition to NFPA 805 and will be submitted separately for NRC approval.

Note there are no NFPA 805 Recovery Actions that require cutting of cables. The Appendix R operator SSA manual actions quoted above from NUREG-0830 Supp. 3 for a Main Control Room evacuation fire event RAI 03 have been superseded by NFPA 805 plant modifications to provide for the capability of isolation / transfer of control to the Primary Control Station, with redundant fusing. These NFPA 805 modifications are included in Attachment S of the LAR. The NFPA 805 Recovery Actions associated with Main Control Room fires are identified and evaluated as VFDRs since they do not occur at the Primary Control Station, RP118B.

August 2011 Page T-6

Ameren Missouri Callaway Plant NFPA 805 Transition Report

1. The Emergency Personnel Hatch is provided for evacuation purposes at El. 2013' as shown on drawing A-2802. The emergency personnel hatch has two bulk head doors SSA on either side of the reactor building wall which are secured by multiple pin latches. RAI The gap between the door and the bulk heads is sealed by double-o-ring gaskets. 04 The bulk heads and hatch doors are in series and provide redundant fire barrier protection. In Modes 1 through 4 the doors are mechanically interlocked to ensure that one door cannot be opened unless the second door is closed.

The emergency personnel hatch opens to fire area RB-1 on the reactor building side and the yard fire area YD-1 on the outside. In the YD-1 fire area the emergency hatch opens into an enclosed stairwell (Room 2202) leading to the outside grade elevation that is separated by a 3-hour barrier from the Reactor Building and contains no fixed ignition sources or equipment. On the RB-1 side the area surrounding the hatch is maintained free of equipment obstructions and combustibles to ensure emergency access to the hatch is maintained. The emergency hatch is robustly designed to meet ASME Section III criteria and there are no significant ignition sources or combustibles on either side of the hatch that could challenge the non-rated hatch.

2. The Equipment Hatch opens to the Yard fire area outdoors and is located on the refueling floor elevation 2047'. The equipment hatch is designed to ASME section III SSA requirements consisting of a welded steel assembly with a double gasketed, flanged, RAI and bolted cover and provided with a moveable concrete missile shield on the 04 outside of the Reactor Building. The equipment hatch opens to fire area RB-1 on the reactor building side and the yard fire area YD-1 on the outside. On the YD-1 side the equipment hatch access platform is 47 feet above grade and is only accessible by stairs or an equipment elevator. There are no fixed combustibles on the platform.

In the RB-1 side the equipment hatch area is maintained free of fixed equipment by design to allow for equipment passage. The emergency hatch is robustly designed to meet ASME Section III criteria and there are no significant ignition sources or combustibles on either side of the equipment hatch.

There have not been any changes to the equipment or emergency personnel hatches or to the plant configuration surrounding either side of the emergency personnel hatch or the equipment hatch that introduced significant fire hazards that would affect the ability of the hatches to perform their intended fire barrier function.

Request As part of this LAR submittal and approval it is requested that the NRC formally document as prior approval that the Emergency Personnel Hatch and the Equipment Hatch in the Reactor Building/Containment walls are acceptable as installed based on the general text of SER Supplement 3 regarding containment penetrations.

August 2011 Page T-8

Ameren Missouri Callaway Plant NFPA 805 Transition Report By letter dated February 1, 1984, the applicant indicated that the existing fuel tank and all piping are seismic Category I. The fuel oil system is a gravity-feed-type system, therefore, no pressurized sprays will occur as a result of a leak. The floor area adjacent to the dike has floor drains. The day tank is provided with level indication that alarms in the control room if there are more than 3 gallons of leakage.

The applicant considers that the current design of the tank is adequate and, on the basis of the information provided, the staff agrees. If any leaks should occur, they would be promptly detected, and the floor drains would collect the majority of the leakage.

On the basis of its review, the staff concludes that the diesel fuel day tank and dike assembly meets the guidelines in Section C.7.i of BTP CMEB 9.5-1, and is, therefore, acceptable.

Subsequent to the NRC approval it was determined that the actual capacity of the emergency SSA diesel generator day tanks are 600 gallons verse 550 gallons and that the day tank dike RAI capacity is 580 gallons or 97% of the tank capacity verses the 110% that was cited in the 05 analysis and the 100% cited in NUREG-0830, Supplement 3.

The reduction in stated dike capacity is not considered to adversely affect the overall performance of the diesel fuel oil day tank dike system in the event of a leak based on the the following:

1) The existing emergency diesel generator fuel oil day tanks and all piping are designed to seismic Category I. The fuel oil system is a gravity-feed-type system, the day tanks are unpressurized tanks vented to the outdoors via piping equipped with flame arrestors, therefore, no pressurized sprays will occur as a result of a leak.

2) The day tanks are provided with level indication that alarms in the control room if there are more than 3 gallons of leakage.

3) The day tanks dike have a capacity of 97% of the day tank volume and the dike area has a floor drain which drains to a covered 900 gallon floor sump designed for combustible liquids.

4) The floor area adjacent to the day tank dike has floor drains.

5) The area adjacent to the day tanks contains no hot surfaces or ignition sources. Any fuel oil on the general floor area will enter the floor drain system and be routed to the sump. Duplex sump pumps are provided to evacuate the sump. The nearest floor drain is approximately 10 outside of the dike.

6) Operations and Security personnel make tours of the diesel generator rooms during each shift.

7) Diesel generator testing is conducted from the control panel within the emergency diesel generator room. Any leakage occurring during normal operation or testing would be detected by plant personnel.

Request As part of this LAR submittal and transition to NFPA 805, it is requested that the NRC formally document as prior approval the current design configuration of the two emergency diesel SSA generator day tanks. The original NRC approval was granted based on the overall design of the RAI emergency diesel generator fuel oil day tank assembly and did not solely rely on the day tank 05 August 2011 Page T-12

Ameren Missouri Callaway Plant NFPA 805 Transition Report dike capacity. Therefore, the basis for the prior NRC approval and the NRC conclusions made SSA in NUREG-0830, Supplement 3, dated 05/1984 remain valid regarding acceptability of the diesel RAI fuel oil day tank dike system in the A and B Emergency Diesel Generator rooms. 05 August 2011 Page T-13

Ameren Missouri Callaway Plant NFPA 805 Transition Report Prior Approval Clarification Request 6 Current Licensing Basis:

Callaway Plant credits the following allowance regarding shift fire brigade staffing as stated in FSAR Standard Plant (SP) Section 16.12.1 Organization - Unit Staff The Unit organization shall be subject to the following:

b. A site Fire Brigade of at least five members (may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence provided immediate action is taken to fill the required positions.) shall be maintained onsite at all times.

The Fire Brigade shall not include the Shift Manager, and the other members of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency.

LIC-16 Background/Basis:

The two hour grace period was originally a part of the Westinghouse Standard Technical Specifications in Section 6.2.2, Unit Staff. As stated in Section 9.5.1.6 of NRC SER NUREG 0830, The applicant has committed to follow the staff standard technical specifications. The staff finds this acceptable.

In the initial Callaway Plant Technical Specifications NPF-30, Technical Specifications, Section 6.2.2.e contained the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> grace period for fire brigade staffing. In later revisions to the Callaway Plant Technical Specifications the requirements related to fire protection were removed and relocated to the FSAR SP. The allowance for the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> grace period is now located in FSAR SP Section 16.12.1 Organization - Unit Staff as stated above.

Request As part of this LAR submittal and transition to NFPA 805, it is requested that the NRC formally document as a prior approval the allowance for a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> grace period for fire brigade staffing as allowed by the FSAR Section 16.12.1 statement.

August 2011 Page T-16

Attachment V: Revisions to Transition Report Attachment V - Fire PRA Quality

Ameren Missouri Callaway Plant NFPA 805 Transition Report Table V-1 Fire PRA Peer Review - Facts and Observations (F&Os)

F&O# Topic Status Fact/Observation Disposition PRM-B2-1 Fire PRA and Closed Scientech Calculation 17671-015 provides The internal events Peer Review Findings Internal Events the disposition of Internal Events F&Os. were reviewed, and their disposition PRA Existing F&Os were ranked based on their documented in Attachment U to the NFPA fire PRA impact. Category A had the highest 805 Transition LAR for Callaway.

impact. Category B had a lesser impact.

Category C had no impact. This F&O was a SUGGESTION, not a FINDING. It involves the disposition of the Basis for Significance: The Fire PRA is Internal Events PRA GAP items by the Fire based on the internal events PRA. Changes PRA. The Fire PRA classified the GAP items to PRA success criteria will impact the Fire as A, B, or C, based on the potential effect PRA. on Fire PRA results. The A and B items LIC were resolved and incorporated into the Fire 13 PRA while the C items were left for future updates. The C items were considered to have minimal effect on Fire PRA. The Westinghouse Peer Review Team interpreted the GAP Assessment process as a work in progress and suggested the work be re-reviewed and finalized. The licensee interpreted the GAP Assessment process as final at the time of the Peer Review and did not respond to this suggestion.

August 2011 Page V-12

Ameren Missouri Callaway Plant NFPA 805 Transition Report Table V-1 Fire PRA Peer Review - Facts and Observations (F&Os)

F&O# Topic Status Fact/Observation Disposition PRM-B4-1 Initiating Events - Closed Callaway developed no new initiating events PRM-B4-1 requires any new initiating event Document for the Fire PRA. Thus, the self assessment to be modeled in accordance with HLR-IE-A Applicability lists this as not applicable. (Completeness), HLR-IE-B (Grouping) and HLR-IE-C (Frequency) from Part 2 of the Basis for Significance: However, to meet ASME PRA Standard. Compliance with this supporting requirement, a defined basis these HLRs is resolved as follows:

is needed to support the claim of nonapplicability of the requirements. HLR-IE-A (Completeness) is met because the Callaway FPRA evaluated every initiating event considered for the internal events PRA. This evaluation is shown in Table 4-1 of Report Callaway -17671-004 Fire Induced Risk Model. Further justification of completeness is not LIC necessary. 14 HLR-IE-B (Grouping) is not applicable because occurrence of fire events (initiating events and consequential events) are individually identified, based on the cable damage from an individual fire scenario.

Grouping of fire initiators (such as caused by spurious operation) is not done.

HLR-IE-C (Frequency) is not applicable because frequency (or conditional probability) of fire events is determined by the probabilities used for circuit analysis reflecting the cable damage in a scenario.

August 2011 Page V-13

Attachment W: Revisions to Transition Report Attachment W - Fire PRA Insights

Ameren Missouri Callaway Plant NFPA 805 Transition Report W.1 Fire PRA Overall Risk Insights Risk insights were documented as part of the development of the FPRA and are provided in Table W-1. The total plant fire core damage frequency (CDF) and large early release frequency (LERF) was derived using the NUREG/CR-6850 methodology for FPRA development and these risk metrics are useful in identifying the areas of the plant where fire risk is greatest. The risk insights generated were also useful in identifying areas where specific contributors might be mitigated via modification, and in understanding the risk significance of MSO combinations.

Using the definition of significant from the combined ASME/ANS PRA Standard RA-Sa-2009 (for the term significant accident progression sequence) the fire initiating events that sum to 95% of the collective CDF or those whose contribution is more than 1% of the total fire CDF are considered to represent the significant fire scenarios. There are 107 scenarios comprising 90%

of the collective fire CDF at Callaway Plant and 180 scenarios contributing to the top 95%. Of these, only 19 scenarios contribute more than 1% on an individual basis to the collective fire CDF. The scenarios contributing more than 1% of the calculated fire risk on an individual basis are described in Table W-1.

W.2 Risk Change Due to NFPA 805 Transition In accordance with the guidance in Regulatory Position 2.2.4.2 of RG 1.205 Revision 1:

The total increase or decrease in risk associated with the implementation of NFPA 805 for the overall plant should be calculated by summing the risk increases and decreases for each fire area (including any risk increases resulting from previously approved recovery actions). The total risk increase should be consistent with the acceptance guidelines in Regulatory Guide 1.174. Note that the acceptance guidelines of Regulatory Guide 1.174 may require the total CDF, LERF, or both, to evaluate changes where the risk impact exceeds specific guidelines. If the additional risk associated with previously approved recovery actions is greater than the acceptance guidelines in Regulatory Guide 1.174, then the net change in total plant risk incurred by any proposed alternatives to the deterministic criteria in NFPA 805, Chapter 4 (other than the previously approved recovery actions), should be risk-neutral or represent a risk decrease.

Table W-2 provides the risk increases associated with the VFDRs. As allowed by RG 1.205, credit for non-fire related modifications that affect the FPRA results has been calculated to offset PRA the risk increase as demonstrated in Table W-2. It is important to note that the risk reduction is RAI 14 based solely on the scope of fire initiating events. Any additional risk reductions that may result from the internal events PRA have not been included. This change is compared to the total baseline fire risk of ~2E-05/year.

The total change in risk associated with the transition to NFPA 805 results in a small risk increase and the total plant fire risk is below 1E-4 for CDF and 1E-5 for LERF. The total change in risk associated with the transition to NFPA 805 results in a risk increase of 1.96E-06 and 4.11E-08 for CDF and LERF, respectively. The total plant risk is not higher than 1E-4 for CDF or 1E-5 for LERF. Therefore these changes are allowable per RG 1.174.

RG 1.205 also requires the licensee to calculate the additional risk of recovery actions. The development of the Fire Risk Evaluations and data for Table W-2 treated all previously approved recovery actions as new. Thus, the CDF and LERF for all recovery actions are included in the Fire Risk Evaluation results presented in Table W-2.

August 2011 Page W-2

Ameren Missouri Callaway Plant NFPA 805 Transition Report PRA Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF) RAI 15 Contribution Scenario Description Scenario Cumulative Risk Insights CCDP IF1 CDF CLERP LERF 1501-1A NG04C- 20.15% 20.15% Failures include MDAFP "B" 1.12E-02 3.63E-04 4.07E-06 1.22E-04 4.43E-08 NonVent via suction valves spurious close (SC), CCW "B" via EGHV16/54 SC and EFHV52 SC, EDG "B" via EFHV60 spurious open (SO), and all 4 RCP seal injection valves (8351A/B/C/D) SC. The fire damage leaves the plant running on Train "A" with no seal injection available from the NCP. Cutsets are dominated by spurious fire-induced failures of CCW "A",

spurious closure of any one RCP seal injection valve (leading to seal LOCA), and failure to initiate recirc after successful injection.

August 2011 Page W-3

Ameren Missouri Callaway Plant NFPA 805 Transition Report PRA Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF)

RAI 15

Contribution Scenario Description Scenario Cumulative Risk Insights CCDP IF1 CDF CLERP LERF 1501-1 NG04C-Vent 5.14% 25.29% This scenario is dominated 3.72E-02 2.79E-05 1.04E-06 5.95E-04 1.66E-08 by an RCP seal LOCA of 176 gal per minute in one or more pumps with successful ECCS injection, but failures in the ECCS recirculation mode due to a) human errors, b) spurious opening of EGTV0030, c) spurious closure of EFHV0052. The loss of seal cooling is caused by spurious closure of the BBHV8351 valves [fire damage] and spurious closure of the CCW thermal barrier cooling isolation valves due to false signal from EGFT0062. Charging pumps and CCW pump are available, but blockage in the seal injection line and the CCW thermal barrier line isolate seal cooling to all RCPs. After 13 minutes, a 176 gpm LOCA is postulated to occur in each pump.

August 2011 Page W-4

Ameren Missouri Callaway Plant NFPA 805 Transition Report PRA Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF)

RAI 15

Contribution Scenario Description Scenario Cumulative Risk Insights CCDP IF1 CDF CLERP LERF YD-SXFR Startup Xfmr 4.56% 29.85% This scenario involves a large 4.49E-04 2.05E-03 9.20E-07 9.93E-06 2.03E-08 transformer fire in the YARD.

It fails offsite power from the main switchyard to PA01 and PA02. Offsite power is also available to NB01 and NB02 from the COOP line. The dominant pathway to core damage is a loss of RCP seal cooling and a failure to provide RCS makeup in response. AFW is available throughout the sequence.

Contributors to risk are failures of both trains of ESW. The non-safety service water is unavailable due to LOSP. Loss of all ESW causes loss of all ECCS, CCW and the charging pumps. Non-safety charging pump is unavailable due to LOOP. Loss of seal cooling leads to RCP seal LOCA, which cannot be mitigated.

August 2011 Page W-5

Ameren Missouri Callaway Plant NFPA 805 Transition Report PRA Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF)

RAI 15

Contribution Scenario Description Scenario Cumulative Risk Insights CCDP IF1 CDF CLERP LERF 4501-2B H2-Sys 4.17% 34.02% This scenario involves a large 1.57E-03 5.35E-04 8.41E-07 4.12E-05 2.20E-08 turbine hydrogen fire with failure of suppression.

Collateral damage in the turbine building fails all offsite power from the main switchyard to NB01, NB02, PA01 and PA02. Offsite power is available to NB01 and NB02 from the COOP line. The dominant pathway to core damage is through a loss of AFW due to a loss of condensate. Included in this fire is a spurious opening of ADLV0079BB and ADLV0079BA, which drains the CST to minimum tech spec level. At nine hours, CST is empty and non-fire related failures of ESW pumps and room coolers lead to a loss of AFW pump suction. Feed and bleed is similarly failed by the same ESW failures which failed ESW water to the AFW system.

August 2011 Page W-6

Ameren Missouri Callaway Plant NFPA 805 Transition Report PRA Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF)

RAI 15

Contribution Scenario Description Scenario Cumulative Risk Insights CCDP IF1 CDF CLERP LERF C10-8s NG02A-Vent 3.20% 37.22% This scenario is started by a 2.45E-02 2.63E-05 6.45E-07 5.94E-04 1.56E-08 fire in NG02A, which causes significant cable damage in C-10. All Train B safety systems are lost by the fire.

Offsite power to PA01 and PA02 are also failed by the fire. Train A of safety systems is unaffected. Offsite power is available to NB01. Core damage is caused by random failures of Train A safety equipment.

C10-17 RP140 3.15% 40.37% This scenario is started by a 2.28E-02 2.79E-05 6.36E-07 5.51E-04 1.54E-08 fire in RP140, which causes significant cable damage in C-10. All Train B safety systems are lost by the fire.

Offsite power to PA01 and PA02 are also failed by the fire. Train A of safety systems is unaffected. Offsite power is available to NB01. Core damage is caused by random failures of Train A safety equipment.

August 2011 Page W-7

Ameren Missouri Callaway Plant NFPA 805 Transition Report PRA Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF) RAI 15 Contribution Scenario Description Scenario Cumulative Risk Insights CCDP IF1 CDF CLERP LERF C9-12 RP139 3.09% 43.46% This scenario is started by a 2.24E-02 2.79E-05 6.24E-07 5.41E-04 1.51E-08 fire in RP139, which causes significant cable damage in C-9. All Train A safety systems are lost by the fire.

Offsite power to PA01 and PA02 are also failed by the fire. Train B of safety systems is unaffected. Offsite power is available to NB02. Core damage is caused by random failures of Train B safety equipment.

RL015/016e RL15/16-Evac 2.25% 45.71% This scenario is a large fire in 1.20E-01 3.80E-06 4.54E-07 3.37E-03 1.28E-08 control board panels RL015 and RL016 in the main control room. Fire is suppressed before is extends beyond the panel RL015/016, but all equipment controlled from this panel is unavailable.

Safe shutdown is provided by safety train B equipment from the Auxiliary shutdown panel.

Offsite power is available to NB02 from the COOP line through PB05 and NB0214.

Failure to provide safe shutdown from the ASP is attributed to human error and random failures of train B equipment.

August 2011 Page W-8

Ameren Missouri Callaway Plant NFPA 805 Transition Report PRA Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF) RAI 15 Contribution Scenario Description Scenario Cumulative Risk Insights CCDP IF1 CDF CLERP LERF 3801T3 TS#3 2.21% 47.92% This scenario represents a 7.64E-01 5.83E-07 4.45E-07 1.09E-02 6.37E-09 transient fire in the upper cable spreading room [C-22],

which causes limited damage, but creates some high frequency undesired events. Seal injection isolation valves from all four RCPs [BBHV8141 and BBHV8351] are damaged in this fire. Loss of seal cooling is virtually guaranteed.

Spurious opening of ADLV0079BA/BB causes CST drain down to the minimum tech spec water level and requires ESW makeup at 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />. Feed and bleed cooling is unavailable due to fire damage to the PORVs.

4501-3 TB-Cat 2.16% 50.08% This scenario is a 5.60E-02 7.79E-06 4.36E-07 1.73E-03 1.35E-08 catastrophic turbine generator fire which fails all equipment and cables in the Turbine Building, including normal offsite power and offsite power from the COOP.

Random failures of NE01 and NE02 lead to station blackout with no potential credited recovery.

August 2011 Page W-9

Ameren Missouri Callaway Plant NFPA 805 Transition Report PRA Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF)

RAI 15

Contribution Scenario Description Scenario Cumulative Risk Insights CCDP IF1 CDF CLERP LERF 3501T11 TS#11 1.79% 51.87% This scenario represents a 1.29E-01 2.80E-06 3.61E-07 1.19E-02 3.33E-08 transient fire in the lower cable spreading room [C-21],

which causes loss of offsite power to PA01 and PA02 and loss of all train A safety equipment. AFW is available from PAL02 and PAL01B.

Random failures of Train B ESW/CCW and charging system to provide seal cooling leads to RCP seal LOCA and core uncovery.

3801T2 TS#2 1.57% 53.44% This scenario represents a 7.61E-01 4.17E-07 3.17E-07 1.08E-02 4.48E-09 transient fire in the upper cable spreading room [C-22],

which causes limited damage, but creates some high frequency undesired events. Seal injection isolation valves from all four RCP's [BBHV8141 and BBHV8351] are damaged in this fire. Loss of seal cooling is virtually guaranteed.

Spurious opening of ADLV0079BA/BB causes CST drain down to the minimum tech spec water level and requires ESW makeup at 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />. Feed and bleed cooling is unavailable due to fire damage to the PORV.

August 2011 Page W-10

Ameren Missouri Callaway Plant NFPA 805 Transition Report PRA Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF) RAI 15 Contribution Scenario Description Scenario Cumulative Risk Insights CCDP IF1 CDF CLERP LERF 4501-1B LO-Sys 1.48% 54.92% This scenario involves a large 1.57E-03 1.90E-04 2.98E-07 4.12E-05 7.81E-09 turbine lube oil system fire.

Collateral damage in the turbine building fails all offsite power from the main switchyard to NB01, NB02, PA01 and PA02. Offsite power is available to NB01 and NB02 from the COOP line. The dominant pathway to core damage is through a loss of AFW due to a loss of condensate. Included in this fire is a spurious opening of ADLV0079BB and ADLV0079AA, which drains the CST to minimum tech spec level. At nine hours, CST is empty and non-fire related failures of ESW pumps and room coolers lead to a loss of AFW pump suction. Feed and bleed is similarly failed by the same ESW failures which failed ESW water to the AFW system.

August 2011 Page W-11

Ameren Missouri Callaway Plant NFPA 805 Transition Report PRA Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF)

RAI 15

Contribution Scenario Description Scenario Cumulative Risk Insights CCDP IF1 CDF CLERP LERF A29-WR A-29 Whole 1.30% 56.22% The scenario is caused by 2.41E-03 1.09E-04 2.63E-07 2.10E-06 2.29E-10 Room Burnup any/all fires in A-29. Fire modeling was not employed in this room. This scenario fails steam line pressure instrumentation on all steam lines causing spurious opening of SG-ASD's. Fire also fails auxiliary feedwater flow indication on several SG's. Failure of the operator to respond to the loss of instrumentation leads to loss of SG cooling and failure of feed and bleed.

4203-0 TS.4203-T5 1.11% 57.33% This is the large floor-area 1.24E-03 1.81E-04 2.24E-07 2.58E-05 4.65E-09 transient fire in zone 4203.

Notable fire-induced failures include the Normal Charging Pump (PBG04) via failure of bus PB03 due to electical faults. In addition, all three Normal Service Water pumps are failed. There is no fire-induced damage to safety-related equipment or offsite power. Fire risk is driven by random common cause failures of the Essential Service Water pumps to run and the electrical faults that fail the Normal Charging Pump.

August 2011 Page W-12

Ameren Missouri Callaway Plant NFPA 805 Transition Report PRA Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF)

RAI 15

Contribution Scenario Description Scenario Cumulative Risk Insights CCDP IF1 CDF CLERP LERF A13-WR A-13 Whole 1.01% 58.34% This scenario is caused by 8.77E-04 2.33E-04 2.04E-07 7.67E-07 1.78E-10 Room Burnup any/all fires in A-13; fire modeling was not employed in this area. Fire-induced damage includes potentially spuriously open Steam Generator "A", "B", and "D" atmospheric steam dump valves ("B" is not recoverable from the MCR), failure of the Motor-Driven AFW Pump "B" and the Turbine-Driven AFW Pump, and potentially spuriously open Steam Generator Blowdown valves "B" and "C". Due to relatively significant damage to the AFW system, risk is dominated by failure of AFW and subsequent failure of Feed and Bleed.

Note1IgnitionFrequency(IF)includesseverityfactorandprobabilityofnonsuppression,whereapplicable

August 2011 Page W-13

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment W - Table W-2 Fire Area Risk Summary VFDR RAs Fire Risk Eval Fire Area Area Description NFPA 805 Basis Fire Area CDF/LERF (Yes/No) (Yes/No) CDF/LERF A-28 Auxiliary Shutdown Panel 4.2.4.2 8.27E-09 / 1.97E-10 Yes No 8.27E-09 / 1.97E-10 Section A A-29 Auxiliary Feedwater Valve 4.2.4.2 2.63E-07 / 2.29E-10 Yes Yes 1.28E-08 / 1.12E-11 Compartment, SG A&D A-30 Auxiliary Feedwater Valve 4.2.4.2 8.81E-08 / 7.75E-11 Yes Yes 0.00E+00 / 0.00E+00 Compartment, SG B&C A-33 Auxiliary Shutdown Panel 4.2.4.2 1.50E-08 / 1.32E-11 Yes No 3.86E-09 / 3.39E-12 Section B AB-1 Auxiliary Boiler Room 4.2.3.2 / No No N/A PRA C-1 Pipe Space and Tank Area 4.2.4.2 6.96E-12 / 6.84E-14 Yes No 6.96E-12 / 6.84E-14 RAI 04-f C-2 Control Building North Cable 4.2.3.2 / No No N/A Chase C-3 Control Building Cable Chase B 4.2.3.2 / No No N/A C-5 Control Building Access Control 4.2.3.2 2.29E-10 / 2.38E-12 No No N/A Area C-6 Control Building Access Control 4.2.3.2 3.51E-10 / 3.86E-12 No No N/A Area C-7 Control Building North Cable 4.2.4.2 2.69E-09 / 2.19E-11 Yes No 2.69E-09 / 2.19E-11 Chase C-8 Control Building Cable Chase B 4.2.3.2 1.15E-11 / 1.65E-14 No No N/A C-9 Switchgear Room A 4.2.4.2 9.31E-07 / 2.10E-08 Yes No 1.09E-07 / 2.45E-09 August 2011 Page W-16

Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment W - Table W-2 Fire Area Risk Summary VFDR RAs Fire Risk Eval Fire Area Area Description NFPA 805 Basis Fire Area CDF/LERF (Yes/No) (Yes/No) CDF/LERF D-2 Diesel Generator B 4.2.3.2 1.10E-08 / 7.23E-11 No No N/A FB-1 Fuel Handling Building 4.2.4.2 5.39E-08 / 1.99E-11 Yes No 4.56E-08 / 7.98E-12 LDF-1 Laundry Decontamination 4.2.3.2 3.78E-09 / 5.42E-12 No No N/A Facility RB-1 Reactor Building General Area 4.2.4.2 2.36E-07 / 1.93E-09 Yes Yes 2.30E-08 / 6.23E-10 RSB-1 RAM Storage Building 4.2.3.2 / No No N/A RW-1 Radwaste Building 4.2.3.2 3.76E-08 / 5.40E-11 No No N/A TB-1 Turbine Building General Area 4.2.4.2 6.54E-06 / 1.36E-07 Yes No 0.00E+00 / 0.00E+00 UNCT Ultimate Heat Sink Cooling 4.2.3.2 2.73E-09 / 4.61E-13 No No N/A Tower A, El. 2000 UNPH Essential Service Water Pump 4.2.3.2 2.69E-09 / 2.62E-11 No No N/A Room A USCT Ultimate Heat Sink Cooling 4.2.3.2 2.73E-09 / 4.61E-13 No No N/A Tower B, El. 2000 USPH Essential Service Water Pump 4.2.3.2 2.72E-09 / 2.69E-11 No No N/A Room B YD-1 Plant Yard Area El. 2000 4.2.4.2 1.03E-06 / 2.18E-08 Yes No 1.68E-08 / 2.94E-12 PRA Total 2.03E-05 / 3.99E-07 1.96E-06 / 4.11E-08 RAI 04-f August 2011 Page W-19

Attachment X: Revisions to Transition Report Attachment X - Other Requests for Approval

$PHUHQ0LVVRXUL &DOODZD\3ODQW1)3$7UDQVLWLRQ5HSRUW

$SSURYDO5HTXHVW



Approval is requested for the 20 foot separation zones in Fire Areas A-1, A-16, A-27 and C-1 as complying with the criteria of no intervening combustible materials or fire hazards and meeting the deterministic compliance criteria of NFPA 805 section 4.2.3.3(b). 

5HTXLUHPHQWV

Callaway Plant Fire Areas A-1, A-16, A-27, and C-1, credit NFPA Section 4.2.3.3(b) to achieve deterministic compliance. NFPA 805 Section 4.2.3.3(b) requires the following:



³:KHUHUHTXLUHGFDEOHVRUHTXLSPHQWRIUHGXQGDQWVXFFHVVSDWKVRIV\VWHPVQHFHVVDU\

WRDFKLHYHDQGPDLQWDLQWKHQXFOHDUVDIHW\SHUIRUPDQFHFULWHULDDUHORFDWHGZLWKLQWKH

VDPHILUHDUHDRXWVLGHRISULPDU\FRQWDLQPHQWRQHRIWKHIROORZLQJPHDQVRIHQVXULQJ

WKDWDWOHDVWRQHVXFFHVVSDWKLVIUHHRIILUHGDPDJHVKDOOEHSURYLGHG



E 6HSDUDWLRQRIUHTXLUHGFDEOHVDQGHTXLSPHQWRIUHGXQGDQWVXFFHVVSDWKVE\D

KRUL]RQWDOGLVWDQFHRIPRUHWKDQIW P ZLWKQRLQWHUYHQLQJFRPEXVWLEOHPDWHULDOV

RUILUHKD]DUGV,QDGGLWLRQDXWRPDWLFILUHGHWHFWRUVDQGDQDXWRPDWLFILUHVXSSUHVVLRQ

V\VWHPVKDOOEHLQVWDOOHGWKURXJKRXWWKHILUHDUHD'





As noted in NRC Generic Letter (GL) 86-10, 20 feet of separation with absolutely no intervening combustible or fire hazards is not always achievable. Therefore, engineering judgment is necessary to evaluate each specific configuration. GL 86-10 provided the general guidance regarding evaluation of intervening combustibles and fire hazards within a 20 ft separation zone.

LIC-18



%DVLVIRU5HTXHVW

This evaluation will consider in-situ combustibles and ignition sources to determine if they represent intervening combustibles or a fire hazard within the 20 ft zone. As noted in GL 86-10, transient combustibles are not considered as an intervening combustible and they are addressed by the Callaway transient combustible control program. All of the 20 ft separation zones evaluated herein are in the current pre-transition Fire Protection Program, and the 20 ft separation zones are clearly marked on the floor of the respective fire areas with red striping and No Combustible Zone lettering. The raceway cables in the 20 ft separation zone raceways are all IEEE-383 qualified Thermoset type cable. Per the GL 86-10 guidance, cables in conduit and cables in completely enclosed cable trays are not considered intervening combustibles. At Callaway, enclosed cable trays are cable trays with both top and bottom tray covers and the tray cover seams are sealed with a standard fire resistive caulk such as 3M CP 25 (Ref E-2R8900 Sht. 53).

Additionally, small items like equipment tags, signage, labels, fire extinguisher tags, hose rack covers, postings comprised of combustible materials (plastic / paper), gai-tronics handsets, and emergency light units exist in small quantities in the areas. However, they are not considered to exist in sufficient quantity to propagate or spread fire across a 20 ft separation zone therefore these are not considered intervening combustibles or fire hazards.

Fire Area C-1 To meet deterministic separation criteria, Fire Area C-1 Fire Zone 3101 is divided into two safe shutdown analysis areas C-1N and C-1S which are separated by a 20 ft separation zone. As shown in Table 4-3 this zone has automatic detection and automatic suppression and the

$XJXVW 3DJH;

$PHUHQ0LVVRXUL &DOODZD\3ODQW1)3$7UDQVLWLRQ5HSRUW

location of the 20 ft separation zone is shown on plant drawing A-2817. Additionally, the redundant circuits are separated by greater than 20 feet. Fire Area C-1 is a piping chase with primarily piping, valves and conduit. There are no ignition sources as defined by NUREG/CR-6850 located within the 20 ft zone and there are no cable trays within the 20 ft separation zone.

There are emergency light units and several small (<5hp motor) MOVs in the 20 ft separation zone. Based on the plant configuration, the 20 ft separation zone has no intervening combustibles or fire hazards and the 20 ft separation zone in fire area C-1 is considered to meet the criteria of NFPA 805 Section 4.2.3.3(b).



Fire Area A-1 To meet deterministic separation criteria, Fire Zones 1206 and 1207 have a 20-foot separation zone. As shown in Table 4-3 this zone has automatic detection and automatic suppression and the location of the 20 ft separation zone is shown on drawing A-2818. Additionally, the redundant circuits are separated by greater than 20 feet. These zones consist of a pipe chase area on the 1988 elevation of the Auxiliary Building containing primarily piping, valves and conduit. There are no ignition sources as defined by NUREG/CR-6850 located within the 20 ft zone and there are no cable trays within the 20 ft separation zone. There are several small AOVs in the 20 ft separation zone. Based on the plant configuration, the 20 ft separation zone has no intervening combustibles or fire hazards and the 20 ft separation zone in fire area A-1 is considered to meet the criteria of NFPA 805 Section 4.2.3.3(b).

Fire Area A-16 To meet deterministic separation criteria, Fire Area A-16, Fire Zone 1408 is divided into two safe shutdown analysis areas A-16N and A-16S which are separated by a 20 foot separation zone.

LIC-18 As shown in Table 4-3 this zone has partial automatic detection and automatic suppression over portions of the 20 ft separation zone. Additionally, the redundant circuits and equipment are separated by much greater than 20 feet. This fire zone is the general corridor area of the Auxiliary Building 2026 elevation and the separation zone is established to separate the two trains of Component Cooling Water components that are located in the large fire area. The location of the 20 ft separation zone is shown on drawing A-2814. During initial Callaway Plant licensing, the configuration of the 20 ft separation zone was reviewed and approved by the NRC in Callaway Plant SSER 3 and that approval is being carried forward (Ref. Attachment K Licensing Action 008). The approval in SSER 3 was stated as follows; SSER 3 states:

³&RPSRQHQW&RROLQJ3XPSV

7KHFRPSRQHQWFRROLQJZDWHUSXPSVDUHORFDWHGRQWKHIRRWHOHYDWLRQRIWKH

$X[LOLDU\%XLOGLQJ3DUWLDOVSULQNOHUV\VWHPVDUHSURYLGHGIRUWKHFRUULGRUDUHDDURXQGWKH

SXPSVKRZHYHUWKHUHLVDQDUHDEHWZHHQWKHSXPSVWKDWGRHVQRWKDYHVSULQNOHU

SURWHFWLRQDQGWKDWFRQWDLQVLQWHUYHQLQJFRPEXVWLEOHV %23FDEOHWUD\V 7KLV

FRQILJXUDWLRQLVQRWLQDFFRUGDQFHZLWK6HFWLRQ&ERI%73&0(%7KH

DSSOLFDQWE\OHWWHUVGDWHG)HEUXDU\DQGDQGFRPPLWWHGWRSURYLGH

ILUHVWRSVLQWKHLQWHUYHQLQJFDEOHWUD\VDGMDFHQWWRRQHRIWKHVSULQNOHUHG]RQHVE\

2FWREHU%HFDXVHRIWKHQDWXUHDQGFRQILJXUDWLRQRIFRPEXVWLEOHVLQWKLVDUHD

WKHILUHVWRSVZRXOGHIIHFWLYHO\SUHYHQWDILUHIURPVSUHDGLQJWRUHGXQGDQWWUDLQV2QWKH

EDVLVRIWKLVFRPPLWPHQWWKHVWDIIILQGVWKDWWKHSURWHFWLRQSURYLGHGIRUWKHFRPSRQHQW

FRROLQJZDWHUSXPSVPHHWVWKHJXLGHOLQHVLQ6HFWLRQ&ERI%73&0(%DQGLV

WKHUHIRUHDFFHSWDEOH'

The cable tray fire stops consist of enclosed cable trays with a section of RTV foam inside the cable tray as shown on plant drawing E-2R8900-Sheet 65. All cable trays that traverse through

$XJXVW 3DJH;

$PHUHQ0LVVRXUL &DOODZD\3ODQW1)3$7UDQVLWLRQ5HSRUW

the 20 ft separation zone are fully enclosed. There are no ignition sources as defined by NUREG/CR-6850 located within the 20 ft separation zone. Only minor combustibles such as a hose station with a cover, a bank of Halon System Bottles, and a small (<5hp) MOV are in the 20 ft separation zone. The 20 ft separation zone has no intervening combustibles or fire hazards and the 20 ft separation zone in Fire Area A-16 is considered to meet the criteria of NFPA 805 Section 4.2.3.3(b) based on prior approval.

Fire Area A-27 To meet deterministic separation criteria, Fire Zone 1403 has a 20 ft separation zone. As shown in Table 4-3 this zone has automatic detection and automatic suppression and the location of the 20 foot separation zone is shown on drawing A-2814. Additionally, the redundant circuits and equipment are separated by much greater than 20 feet. Fire Zone 1403 contains the Control Rod Drive MG Sets and the Reactor Trip Switchgear along with numerous electrical panels and their associated cable trays. As shown in Figure 1, there are cable trays, a motor control center, an air handling unit, and one of the control rod drive motor generator sets located within the 20 ft separation zone.

The two raceways containing redundant trains of cables are separated by approximately 58 ft.

Figure 1 shows the location of the two redundant raceways relative to the 20 ft separation zone and to each other. One set of redundant raceways are cable trays that enter the fire area from the south wall, travel approximately 8 ft where the cable trays turn west to exit through a fire rated penetration seal into the Control Building. The bottom trays are approximately 15 ft above the floor. The second redundant raceway is a conduit that crosses the zone at height 18 ft above the floor.

A 480 volt non-safety related motor control center (MCC) PG20G is located in the 20 ft separation zone. As shown in Figure 1, the cables in tray exiting this MCC 1) route to the cable LIC-18 trays that do not cross through the 20 ft separation zone or 2) route to cable trays which are fully enclosed, so they pose no concern as an intervening combustible. Based on the configuration the MCC is not considered a fire hazard that can result in a fire that will propagate across the 20 ft separation zone.

Air handling unit SGL20, the air handling unit for the room itself, is located within the 20 ft separation zone. The air handling unit is not located near secondary combustibles. Based on the distance from the cable trays and the insignificant amount of combustibles associated with the unit, the air handling unit is not considered a fire hazard that can result in a fire that will propagate across the 20 ft separation zone.

Control Rod Drive Motor Generator set SF01 is located within the 20 ft separation zone but located greater than 17 feet from the nearest redundant train raceway. The motor generator set has exposed cables in tray that route to MCC SF103A which is also located in the 20 ft separation zone. Figure 1 shows the location of these two components. There are no exposed secondary combustibles near the motor generator set or its MCC that if ignited can propagate fire across the 20 ft separation zone. Based on the configuration the MCC and SF01 are not intervening combustibles or a fire hazard that can result in a fire that will propagate across the 20 ft separation zone.

As shown in Figure 1 there are two sets of non-safety related cable trays that traverse the 20 ft separation zone. As previously discussed, all the exposed cables are Thermoset type IEEE-383 qualified cables. The set of cable trays on the west side of the 20 ft separation zone are located approximately 2'-6 from the west wall and contain three horizontally stacked cable

$XJXVW 3DJH;

$PHUHQ0LVVRXUL &DOODZD\3ODQW1)3$7UDQVLWLRQ5HSRUW

trays. These cable trays are fully enclosed for over 50 ft of the tray length to a point beyond the redundant conduit raceway location. The enclosed portion of the cable trays extends into the 20 ft separation zone approximately 8 ft. The set of three cable trays on the east side of the 20 ft separation zone are fully enclosed for greater than 50 ft of the tray length to a point beyond the redundant conduit raceway location. The enclosed portion of the cable trays extend into the 20 ft separation zone for approximately 4 feet and the cable trays turn and exit the fire area and do not cross through the 20 ft separation zone.

In the A-27 fire area, the 20 ft separation zone has fixed ignition sources and combustibles in the form of exposed cable. However, the separation of required cables and equipment of redundant success paths is by a horizontal distance of approximately 58 ft and automatic fire detectors and an automatic fire suppression system is installed throughout the fire area.

Horizontal stacks of cable trays are fully enclosed for significant portions of their travel path in the 20 ft separation zones thereby eliminating the possibility a fire can propagate across the 20 ft separation zone via the cable trays. The fixed ignition sources that exist in the 20 ft separation zone are located such that their ignition will not result in a fire that can propagate across the 20 ft separation zone. There are no intervening combustibles or fire hazards in the 20 ft separation zone therefore, the 20 ft separation zone in Fire Area A-27 is considered to meet the criteria of NFPA 805 Section 4.2.3.3(b).

LIC-18











$XJXVW 3DJH;

$PHUHQ0LVVRXUL &DOODZD\3ODQW1)3$7UDQVLWLRQ5HSRUW

1XFOHDU6DIHW\DQG5DGLRORJLFDO5HOHDVH3HUIRUPDQFH&ULWHULD

Based on the analysis presented above, Fire Areas A-1, A-16, A-27 and C-1 all comply with the criterion of NFPA 805 Section 4.2.3.3. The evaluation shows that there is adequate separation along with automatic detection and suppression to ensure that one train of the redundant trains of circuits that exist within each of the fire areas remains free of fire damage. The design meets deterministic criteria therefore this is not a performance based approval request.

6DIHW\0DUJLQDQG'HIHQVHLQ'HSWK

Based on the analysis presented above, Fire Areas A-1, A-16, A-27 and C-1 all comply with the LIC-18 criterion of NFPA 805 Section 4.2.3.3. The evaluation shows that there is adequate separation along with automatic detection and suppression to ensure that one train of the redundant trains of circuits that exist within each of the fire areas remains free of fire damage. The design meets deterministic criteria therefore this is not a performance based approval request.

&RQFOXVLRQ

NRC approval is requested to document that the 20 ft separation zones within Fire Areas A-1, A-16, A-27 and C-1 as described and evaluated above comply with the criteria of no intervening combustible materials or fire hazards and meet the deterministic compliance criteria of NFPA 805 Section 4.2.3.3(b). 

$XJXVW 3DJH;

to ULNRC-05876 Enclosure 2, Computer Input Files for Fire Modeling RAIs 2b, 3b and 3t The following files are provided on CD:

• FDS computer input file requested for Fire Modeling RAI 2b

• CFAST computer input files requested for Fire Modeling RAI 3b

• FDS computer input files requested for Fire Modeling RAI 3t Page 1 of 1