ML12108A240

Enclosure 1 to ULNRC-05851, Request for Additional Information (RAI) with Callaway Plant Response

ML12108A240
Person / Time
Site:	Callaway
Issue date:	04/17/2012
From:	Ameren Missouri
To:	Office of Nuclear Reactor Regulation
Shared Package
ML121080490	List: ML12108A240 ULNRC-05851, Response to Request for Additional Information Regarding Adoption of National Fire Protection Association Standard NFPA 805
References
TAC ME7046, ULNRC-05851
Download: ML12108A240 (158)

Text

Enclosure 1 to ULNRC-05851 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: Not used.

Attachment E: Not used.

Attachment F: Not used.

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

Attachment I: Not used.

Attachment J: Not used.

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.

Page 1 of 96 to ULNRC-05851 Attachment P: Not used.

Attachment Q: Not used.

Attachment R: Not used.

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

Attachment V: Not used.

Attachment W: Revisions to the Transition Report Attachment W - Fire PRA Insights Page 2 of 96 to ULNRC-05851 Section 1: Response to Fire Modeling RAIs Fire Modeling RAI 01 Section 2.7.3.2, "Verification and Validation," of NFPA 80S 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."

x Flame Height (Method of Heskestad) x Plume Centerline Temperature (Method of Heskestad) x Radiant Heat Flux (Point Source Method) x Hot Gas Layer (Method of MQH) x Hot Gas Layer (Method of Beyler) x Hot Gas Layer (Method of Foote, Pagni, and Alvares [FPA])

x Hot Gas Layer (Method of Deal and Beyler) x Ceiling Jet Temperature (Method of Alpert) x Smoke Detection Actuation Correlation (Method of Heskestad and Delichatsios)

Page 3 of 96 to ULNRC-05851 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 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.

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:

x Heat Detection Actuation Correlation x Sprinkler Activation Correlation x Corner and Wall Heat Release Rate x Correlation for Heat Release Rates of Cables (Method of Lee) x 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.

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.

x Hot Gas Layer (HGL) Calculations using FDS x Sprinkler Actuation Calculation using FDS x Temperature Sensitive Equipment Zone of Influence Study using FDS x 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.

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.

x HGL Calculations using CFAST (Version 6)

Page 4 of 96 to ULNRC-05851 x Temperature Sensitive Equipment Hot Gas Layer Study using CFAST x Control Room Abandonment Calculation using CFAST 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.

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.

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.

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.

Page 5 of 96 to ULNRC-05851 Response to Fire Modeling RAI 01

a. Both the Fire Modeling Database (FMDB) and the Transient Analysis Worksheet were verified, by black box testing, to ensure that the results were identical to the verified and validated models. Black box testing (also called functional testing) is testing that ignores the internal mechanism of a system or component and focuses solely on the outputs generated in response to selected inputs and execution conditions.

The process compared results from the FMDB and Transient Analysis Worksheets to those produced by the NUREG-1805 FDTs and FIVE when identical inputs were entered into both.

Since the correlations from NUREG-1805 FDTs and FIVE-Rev. 1 were verified and validated in NUREG-1824, and the results match the results produced by the FMDB and the Transient Analysis Worksheet, by the transitive property, the FMDB and the Transient Analysis Worksheet is verified and validated with respect to NUREG-1824.

The results of this verification are documented in R1984-001-002, Verification and Validation of Fire Modeling Tools and Approaches for Use in NFPA 805 and Fire PRA Applications.

b. This RAI response will be provided with supplemental correspondence.

c. This RAI response will be provided with supplemental correspondence.

d. This RAI response will be provided with supplemental correspondence.

e. This RAI response will be provided with supplemental correspondence.

f. This RAI response will be provided with supplemental correspondence.

g. This RAI response will be provided with supplemental correspondence.

h. The Multi-Compartment Hot Gas Layer Analysis is not a fire model. This methodology refers to task input to the Multi-Compartment Analysis from the Detailed Fire Modeling task. The detailed fire modeling determined hot gas layer through the following methods: McCaffrey, Quintiere and Harkleroad (MQH); Beyler, Foote, Pagni and Alvares; Deal and Beyler; Fire Dynamics Simulator (FDS); and Consolidated Model of Fire Growth and Smoke Transport (CFAST). These models are detailed elsewhere in Attachment J to the Transition Report.

Since the Multi-Compartment Analysis Hot Gas Layer Analysis is not a fire model and the methods used to determine hot gas layer are correctly identified and documented elsewhere in Attachment J, the reference to Multi-Compartment Analysis Hot Gas Layer Analysis is being deleted from section 4.5.1.2 of the Transition Report.

i. This RAI response will be provided with supplemental correspondence.

j. This RAI response will be provided with supplemental correspondence.

Page 6 of 96 to ULNRC-05851 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 ..."

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.

Response to Fire Modeling RAI 02

a. Fire modeling 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. 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. 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 area-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 Page 7 of 96 to ULNRC-05851 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:

Aleatory uncertainties identified within the fire modeling parameters include:

x Detector response reliability and availability x Automatic suppression system reliability and availability x Manual suppression reliability with respect to time available Epistemic uncertainties which impact the zone of influence and time to damage range include:

x Heat release rates (peak HRR, time to reach peak, steady burning time, decay time) x Number of cabinet cable bundles x Ignition source fire diameter x Room ventilation conditions x Sprinkler Response Time Index (RTI), C factor, and activation temperature x Detector activation temperature, geometry and obscuration activation x Soot yield x Fire growth assumptions (cable tray empirical rule set, barrier delay) x Cable fire spread characteristics for horizontal and vertical trays x Transient fires (peak HRR, time to reach peak, location factor, detection time) x Oil fires (spill assumptions) x Assumed target location x Target damage threshold criteria x Manual detection time x Mean prompt suppression rate x Manual suppression rate x Welding and cutting target damage set x 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.

Page 8 of 96 to ULNRC-05851 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, the safety margin should be sufficient 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:

x 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.

x 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.

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

x 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.

x For transient fire impacts, a large bounding transient zone assumes all targets within its Zone of Influence (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 smaller zone of influence and varying damage times. This approach is implemented to minimize the multitude of transient scenarios to be analyzed.

x 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.

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

x 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.

x 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.

x 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.

Page 9 of 96 to ULNRC-05851 x 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.

x 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.

x 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).

Attachment J of the Callaway Plant NFPA 805 Transition Report makes the statement that the predictions are deemed to be within the bounds of experimental uncertainty for the following fire model calculations:

x Hot gas layer height and temperature using Fire Dynamics Simulator (FDS) x Hot gas layer height and temperature using the Consolidated Fire Growth and Smoke Transport Model (CFAST) x Radiant heat using FDS This statement is provided to show that the models are within or very near experimental uncertainty, as determined by NUREG-1824, Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications, Final Report, dated April 2007. This statement is further substantiated as follows:

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.

Page 10 of 96 to ULNRC-05851 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 on The 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 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.

Radiant Heat using FDS The predictive capability of this parameter in 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.

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 Page 11 of 96 to ULNRC-05851 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.

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. This RAI response will be provided with supplemental correspondence.

Page 12 of 96 to ULNRC-05851 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. 0, 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.

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 were 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 (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.

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 Page 13 of 96 to ULNRC-05851 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.

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.

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.

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.

j. During the audit, the NRC staff noted that Section C21.2 of fire modeling report R1984-001-001 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 Page 14 of 96 to ULNRC-05851 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.

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 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).

l. During the audit, the NRC staff noted that Section C21.3.5.1 of fire modeling report R1984-001-001 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.

o. During the audit, the NRC staff noted that Section C21.5 of fire modeling report R1984-001-001 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.

Page 15 of 96 to ULNRC-05851 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.

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 multi cabinet 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 17671-010b, 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.

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 8).

Response to Fire Modeling RAI 03

a. This RAI response will be provided with supplemental correspondence.

b. This RAI response will be provided with supplemental correspondence.

c. This RAI response will be provided with supplemental correspondence.

d. Applying the FDS analysis of a single fire compartment (i.e., Fire Compartment C-31) to additional fire compartments (i.e., Fire Compartments A-11 and C-30), is justified when the FDS analysis is developed with conservative parameters that bound the results for the other compartments.

The time to sprinkler activation in a fire zone or compartment, determined using FDS, may be applied to another fire zone/compartment given the following conditions:

Page 16 of 96 to ULNRC-05851

• Fire compartment volume is smaller than or equivalent to what was used in the FDS analysis

• The radial and vertical distance from the detector to the fire is less than or equal to the FDS analysis

• There is no fire size smaller than that used in the FDS analysis that could result in the Fire Damage State being analyzed (e.g. results in hot gas layer)

• Ventilation conditions are the same

• Sprinkler/detector type and properties are the same

• Compartment boundary material (e.g. concrete) is the same Fire Compartments A-11, C-30 and C-31 meet the conditions outlined above. The applicability of each parameter is discussed in detail in R1984-001-001, Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling Revision 0 to justify the use of a single FDS analysis to bound the multiple configurations. The justification includes how the ignition source location and secondary combustibles are considered in the analysis.

e. This RAI response will be provided with supplemental correspondence.

f. This RAI response will be provided with supplemental correspondence.

g. Sprinkler water spray characteristics were not used. FDS is used only to determine timing to sprinkler activation. A graphical output from Smokeview, depicting discharge spray is included only for demonstrative purposes.

h. This RAI response will be provided with supplemental correspondence.

i. This RAI response will be provided with supplemental correspondence.

j. This RAI response will be provided with supplemental correspondence.

k. This RAI response will be provided with supplemental correspondence.

l. This RAI response will be provided with supplemental correspondence.

m. This RAI response will be provided with supplemental correspondence.

n. This RAI response will be provided with supplemental correspondence.

o. Contrary to the statement in Section C21.5 of fire modeling report R1984-001-001, Fire Compartment C-21 has a ceiling height of 15 feet and this dimension was used in the FDS simulation. The report has been corrected to reflect the 15 foot ceiling height. Fire Compartment C-22 has a ceiling height of 12 feet.

Page 17 of 96 to ULNRC-05851 It was not necessary to model Fire Compartment C-22 separately because the analysis developed for C-21 produces bounding suppression and detection activation times for C-22.

The fire location in the modeled configuration was chosen to represent a worst case placement of the fire with respect to smoke detectors and sprinkler heads, which delayed activation.

The detectors and sprinklers in both compartments are located at or near ceiling level. The smoke detectors and sprinklers installed in Fire Compartment C-21, with a 15 foot ceiling, are therefore at a greater vertical distance from the floor-based fires than would occur in C-22, where the ceiling is only 12 feet. If the scenario had been modeled for C-22, the difference in vertical distance from the fire source would result in shorter times to detector and sprinkler activation.

The fire was also placed at a location that bounds the maximum radial distance from the smoke detectors and sprinklers for both compartments. In addition, the fire was placed in a beam pocket that did not have installed smoke detection. Using the maximum radial distance, and including the beam pocket configuration, ensures that the most conservative configuration was modeled (i.e., the resultant timing estimations are based on the most delayed time to heating of the sprinklers and the most delayed time for combustion products to reach the smoke detectors).

Other than ceiling height, the two compartments have similar geometry and ventilation conditions, have the same type of sprinklers and detectors installed, and are constructed of the same boundary materials. Since the fire was placed at a maximum vertical and radial distance from the smoke detectors and sprinklers, and since the ceiling height for C-22 is less than C-21, the detector activation times generated in Fire Compartment C-21 bound the results of C-22.

p. This RAI response will be provided with supplemental correspondence.

q. This RAI response will be provided with supplemental correspondence.

r. This RAI response will be provided with supplemental correspondence.

s. This RAI response will be provided with supplemental correspondence.

t. This RAI response will be provided with supplemental correspondence.

Page 18 of 96 to ULNRC-05851 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 The correct compliance statement for Section 3.3.7.1 is "complies with required action". The LAR Transition Report Attachment A, NEI 04-02 Table B Transition of Fundamental Fire Protection Program and Design Elements, has been updated with this correction as shown in Attachment A to this Enclosure.

Page 19 of 96 to ULNRC-05851 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 The correct compliance statement for Section 3.4.1(a)(1) is "complies with required action". The updated LAR Transition Report Attachment A, NEI 04-02 Table B Transition of Fundamental Fire Protection Program and Design Elements, with this correction is provided in Attachment A to this Enclosure.

Page 20 of 96 to ULNRC-05851 Fire Protection Engineering RAI 03 In Attachment A of the LAR, NEI 04-02 Table B Transition of Fundamental Fire Protection Program and Design Elements, 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 The correct compliance statement for section 3.4.2 is "complies with required action". The updated LAR Transition Report Attachment A, NEI 04-02 Table B Transition of Fundamental Fire Protection Program and Design Elements, with this correction is provided in Attachment A to this Enclosure.

Page 21 of 96 to ULNRC-05851 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 LAR Transition Report Attachment A, NEI 04-02 Table B Transition of Fundamental Fire Protection Program and Design Elements, Section 3.4.2.3 has been revised and the required action has been added to the section. Section 3.4.2.4 has been revised and the correct compliance statement is Complies with Clarification. Section 3.4.2.4 does not have required action. The revised LAR Transition Report Attachment A with this correction is provided in Attachment A to this Enclosure.

Page 22 of 96 to ULNRC-05851 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 The correct compliance statement for Section 3.4.4 is "complies with required action". The updated LAR Transition Report Attachment A, NEI 04-02 Table B Transition of Fundamental Fire Protection Program and Design Elements, is provided in Attachment A to this Enclosure.

Page 23 of 96 to ULNRC-05851 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 The compliance basis statement for Section 3.5.15 has been updated to identify that each mobile unit has equipment equivalent to that of three hose houses. The updated LAR Transition Report Attachment A, NEI 04-02 Table B Transition of Fundamental Fire Protection Program and Design Elements, is provided in Attachment A to this Enclosure.

Page 24 of 96 to ULNRC-05851 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 Section 3.6.2 has two compliance statements; one is complies with clarification which is applicable to all hose stations except those protecting the ESW pump house and the clarification is provided to identify this exception. The second compliance statement applicable to the hose stations protecting the ESW pump house is Submit for NRC Approval. The updated LAR Transition Report Attachment A, NEI 04-02 Table B Transition of Fundamental Fire Protection Program and Design Elements, with enhanced wording is provided in Attachment A to this Enclosure.

Page 25 of 96 to ULNRC-05851 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 This RAI response will be provided with supplemental correspondence.

Page 26 of 96 to ULNRC-05851 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 The correct compliance statement for Section 3.9.3 and Section 3.10.2 is "complies". The updated LAR Transition Report Attachment A, NEI 04-02 Table B Transition of Fundamental Fire Protection Program and Design Elements, with this correction is provided in Attachment A to this Enclosure.

Page 27 of 96 to ULNRC-05851 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 Callaway Plant corrective action request, CAR 201110797, documents the incident described above.

While performing a monthly surveillance run of the 'B' Emergency Diesel Generator (EDG) per OSP-NE-0001B, Standby Diesel Generator B Periodic Tests, a problem occurred that resulted in the jacket water heater remaining energized and eventually resulting in the pipe insulation and coating to catch on fire. Troubleshooting activities revealed that a screw had prevented the starter for the feeder breaker (NG04DEF4) to the jacket water heater from de-energizing the heater.

The design of the jacket water keep warm system, which includes the jacket water keep warm pump and heater, is to cease operation once the low speed relay is made up during the starting sequence and then subsequently start once the EDG is secured. During this event the keep warm pump was successfully stopped, thus stopping water flow through the still energized heater. As the heater remained energized, the lack of flow allowed the water inside the heater to boil. Once all or the majority of the water in the heater had boiled off, the 42 kW heater continued to heat the surrounding enclosure which eventually lead to the insulation and paint catching on fire.

The small fire initiated by the heater failure was quickly extinguished by the Operations personnel performing the 'B' EDG test. However, should the failure of the heater have occurred during an actual emergency demand of the 'B' EDG, Operators would not have necessarily been present in the room to extinguish the fire. A review of the event revealed that the small fire was at the center of the heater and was caused by burning paint and insulation on the heater. Other than the flexible conduit for the heater cables, no other combustibles were present in the area of the fire. This flexible conduit could have been exposed to flames during the fire. However, as evident from post fire inspection the conduit did not exhibit signs of fire damage, indicating that the intensity of the fire was low. Due to the size and limited combustibles in the area, fire spread would not have been likely. The fire detection system in the room was not activated during the event. Had the fire detection system been activated by the thermal detectors the suppression system would have been charged. However, there was insufficient heat present to melt the sprinkler heads and discharge the fire suppression system. Therefore, it can be reasonably assumed that without operator intervention the heater fire would have consumed the remaining combustible material (insulation and paint on the heater) and self-extinguished without adversely impacting the function of the 'B' EDG.

This incident has no direct impact on the NFPA 805 License Amendment Request or transition process and no actions or document revisions were necessary as a result of this internal OE event. The normal Page 28 of 96 to ULNRC-05851 update process for the FPRA will consider the impact of the event on EDG reliability/availability assumed within the analysis.

Page 29 of 96 to ULNRC-05851 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 Callaway Plant has reviewed LAR Transition Report Attachment A, Section 3.5.1 and Attachment L, Approval Request 1, regarding the fire protection water supply in response to this RAI. As a result of the review, the compliance basis for LAR Transition Report Attachment A, Section 3.5.1a, has been revised to credit the water supply criteria of Section 3.5.1(b) which is based on a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> flow rate versus the fixed storage capacity of two 300,000 gallon tanks of Section 3.5.1(a) cited in the submittal.

The Callaway Plant fire protection water supply current licensing basis uses the criteria of providing a two hour water supply so this change carries forward the current design / licensing basis. The revised LAR Transition Report Attachment A is attached as Attachment A to this Enclosure.

Site fire protection procedure APA-ZZ-00703, Fire Protection Operability Criteria and Surveillance Requirements, requires that the two fire water tanks maintain a volume of 260,000 gallons (tank water level 31 feet) each to remain functional. The tanks are normally maintained at a nominal water level of 34 feet. Each tank is provided with local level indication and an automatic make up capability of 650 gpm that initiates when the tank level reaches a water level of 33.6 feet. The local tank level is verified by Operations department personnel during normal daily rounds.

Due to the design of the automatic make up system, it does not automatically makeup when the fire water pumps are being used in the configuration requested in the LAR Transition Report Attachment L, Approval Request 1. Therefore, to ensure that the non-fire protection water use does not affect the tank functions, additional monitoring requirements are added which include monitoring the tank level of both fire water storage tanks to ensure tank levels are maintained above water level 31 feet. The revised LAR Transition Report Attachment L, Approval Request 1 is provided as Attachment L of this Enclosure.

Page 30 of 96 to ULNRC-05851 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 As a result of the Callaway Plant review and response to Fire Protection Engineering RAI 11, the compliance basis statement contained in the LAR Transition Report Attachment A, Section 3.5.1, has been revised to credit the criteria of Section 3.5.1(b) as the basis for the Callaway Plant fire protection water supply. LAR Transition Report Attachment L, Approval Request 1 is therefore correct as submitted and consistent with the revised LAR Transition Report Attachment A, Section 3.5.1 response provided as Attachment A to this Enclosure.

Page 31 of 96 to ULNRC-05851 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 This RAI response will be provided with supplemental correspondence.

Page 32 of 96 to ULNRC-05851 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 The condition has been entered in the site corrective action program and compensatory measures have been established per the fire protection program. LAR Transition Report Attachment S, Table S-1 has been revised to remove reference to Modification 201002877. Additionally, Attachment S, Table S-2 has been revised to add plant modification MP 12-0009 as an implementation item. MP 12-0009 has also been added as an implementation item to LAR Transition Report Attachment A, Table B-1 Section 3.9.1(1). At the completion of the modification, the sprinkler systems within the affected fire areas will be in compliance with the requirements of NFPA 13 1976.

The revised LAR Transition Report Attachment S, Table S-1 & Table S-2 are provided as Attachment S of this Enclosure. The revised LAR Transition Report Attachment A, Table B-1 is provided in Attachment A to this Enclosure.

Page 33 of 96 to ULNRC-05851 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 It is noted the referenced LAR Transition Report Attachment A, section is applicable to sprinkler systems, therefore, this response will be provided specific to LAR Transition Report Attachment A Section 3.6.1.

LAR Transition Report Attachment A, Section 3.6.1 identifies that the Callaway Plant standpipe and hose stations are designed and installed in accordance with the NFPA 14, 1976 requirements for Class II service as previously approved by the NRC. The Callaway Plant compliance with specific sections of NFPA 14-1976 is contained within Callaway Plant Calculation KC-27, NFPA Code Conformance Review. As noted in the RAI, the Callaway Plant fire protection system normal working pressure may range as high as 150 to 160 psi for the standpipes. Specific to NFPA 14, Section 4-4.2 Callaway Plant identified that we Comply with Clarification stating that The pressure reducing devices on the hose valves have been removed, that the plant fire brigade is trained in the use of high pressure hose and hose use is restricted. Since the hose use is restricted, the intent of the code is met and the configuration is acceptable. NFPA 14, Section 4-4.2 is intended as a safety precaution to ensure that in cases where standpipe pressures are high, that an untrained individual does not use the hose and get injured. At the Callaway Plant, the fire brigade is trained in the use of high pressure hose and are the only persons who will use the hose stations for firefighting.

As noted in calculation KC-27, the hose station valves were originally supplied with pressure reducing devices. During an NRC inspection documented in NRC Inspection Report 483/87018 the NRC issued the following Unresolved Item (URI).

"The once per three year surveillance check of the fire hose stations (TS 4.7.10.4) to determine operability by partially opening each hose station valve was discussed as related to the fire hose pressure reducing devices (PRD's). During plant tours the inspector observed the installation of PRD's on each of the fire hose stations. The Licensee was unable to provide documentation to demonstrate that the as-installed setting of these PRD's was adequate.

Section E.3(d) of the SNUUPS response to the Auxiliary Power Conversion and Systems Branch 9.5-1, Appendix A Section, revision 15, dated June, 1984 specifies that all hose stations are equipped with PRD's where required by code. Paragraph 2-1.3.3 of the National Fire Page 34 of 96 to ULNRC-05851 Codes, pamphlet 14-1978 indicates that a pressure regulating device is arranged to regulate pressure at the hose valve outlet to a pressure not exceeding 100 psi.

According to the Licensee, no pre-operational test was performed to determine the appropriate setting of the PRD's, however; the Safety Supervisor indicated that the fire brigade was trained to properly handle standpipe hose pressures found in the plant. Previous regional discussions with NRR have identified the acceptability of certain criteria in lieu of strict conformance with the code as follows:

To have the fire brigade training program require that a minimum of two personnel man the fire hose with only those personnel trained on fire hose use allowed to use it.

Fire Brigade members need to be trained in the use of fire hose stations up to the maximum pressure found on these hose stations.

Post caution or advisory type signs on any standpipe(s) having outlet greater than 100 psi.

This is considered an Unresolved Item (483/87018-03) pending further evaluation by the licensee and review of that evaluation by the NRC in determining fire brigade and or equipment readiness to properly control fire hose station pressures in excess of that discussed in the code."

To address the URI issue, Callaway Plant implemented modification CMP 88-1009a which accomplished the following; x Removed the hose rack valve pressure reducing devices, x The fire brigade training program was revised to require that a minimum of two qualified personnel man the fire hose, x Ensured fire brigade members were trained in the use of fire hose stations up to the maximum pressures found on these stations standpipes, and x Posted caution or advisory type signs on any standpipes having outlet pressures greater than 100 psi.

The URI was closed in a subsequent fire protection inspection as noted in NRC Inspection Report 883/94-005 accepting the actions taken as noted below.

(Closed) Unresolved Item (483/87018-03) "Fire Hose Station Pressure" During an inspection conducted on June 1 through June 5 and on June 29, 1987, the licensee was unable to provide the NRC inspector with documentation concerning the setpoint of the pressure reducing devices (PRD's) on the fire hose stations. Pending further evaluation by the licensee and review of that evaluation by the NRC in determining fire brigade readiness to properly control fire hose station pressure the unresolved item was issued.

The licensee committed to:

Page 35 of 96 to ULNRC-05851 x Train the fire brigade members in the safe handling of high pressure fire hoses and in the use of multiple personnel when handling high pressure fire hoses.

x Label fire hose stations warning personnel of the potential that the hose station pressure could be up to 150 psig.

The inspectors verified through plant walk downs that the warning signs were in place.

Through interviews with random members of the plant fire brigade, the inspectors determined that the brigade members were aware of the hazards associated with high pressure fire hoses and had been trained how to handle those hazards.

This unresolved issue is closed.

To provide clarification of the basis for compliance with NFPA 14, Section 4-4.2, Calculation KC-27 has been updated to include the discussion above. The conclusions for compliance with NFPA 14-1976 section 4-4.2 stated in Calculation KC-27 remain valid and in cases where standpipe pressures exceed 100 psi, the actions taken meet the intent of NFPA 14 section 4-4.2.

Page 36 of 96 to ULNRC-05851 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.

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 Page 37 of 96 to ULNRC-05851 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 This RAI response will be provided with supplemental correspondence.

Page 38 of 96 to ULNRC-05851 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 01 Callaway Plant has performed a comparison or gap analysis between NEI 00-01, Revision 1, and NEI 00-01, Revision 2. Based on the gap analysis there are no significant differences between alignment with NEI 00-01, Revision 1 and NEI 00-01, Revision 2 for the Callaway Plant. LAR Transition Report Section 4.2.1.1 is revised to provide a discussion of these results. The full gap analysis has been added to Callaway Plant Calculation KC-26, Nuclear Safety Capability Assessment, The revised LAR Transition Report Section 4.2.1.1 is provided in Attachment 1 to this Enclosure.

Page 39 of 96 to ULNRC-05851 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 02 The Instrument Air System has not been credited or analyzed in the Callaway Plant NSCA and NPO.

Instrument air system pressure is assumed to exist if it can have an adverse consequence (i.e., air 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 de-energized). 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). To clarify this position, LAR Transition Report Attachment B, has been revised to provide the above clarifying text. The revised LAR Transition Report Attachment B text is provided in Attachment B to this Enclosure.

Page 40 of 96 to ULNRC-05851 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 03 The text that is referenced within the RAI is text from NUREG-0830, Safety Evaluation Report related to the operation of Callaway Plant, Unit No. 1, Supplement 3, dated 05/1984. There are no NFPA 805 Recovery Actions that require cutting of control power cable. NFPA 805 required plant modifications provide for the capability to isolate and transfer control of the fire affected component to the local control station, with redundant fusing. These NFPA 805 modifications are included in Attachment S of the Transition Report included with the LAR. 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. LAR Transition Report Attachments B and Attachment T, Clarification Request 1 have been revised to clarify that those operations are no longer required. The revised portions of the LAR Transition Report Attachment B and T are provided as Attachments B and T to this Enclosure.

Page 41 of 96 to ULNRC-05851 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 04 The LAR Transition Report Attachment T, Clarification Request 2 has been revised to identify that there are no significant ignition sources or combustible loading in the vicinity of the emergency personnel hatch or equipment hatch that can challenge the non-rated penetrations. Additionally, the request is revised to identify that there have been no significant changes to the areas surrounding the hatches since original NRC approval. The revised LAR Transition Report Attachment T is provided in Attachment T to this Enclosure.

Page 42 of 96 to ULNRC-05851 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-05 The LAR Transition Report Attachment T, Clarification Request 4 has been revised to add the diesel fuel oil day tank dike system capacity and additional discussion is provided regarding the system adequacy with the reduced capacity of less than 100 percent. The revised LAR Transition Report Attachment T is provided as Attachment T to this Enclosure.

Page 43 of 96 to ULNRC-05851 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 06 The components and/or cables associated with the actions to recharge the nitrogen accumulators are included in the NSCA equipment list and the actions have been evaluated and demonstrated to be feasible. A discussion indicating that the qualitative risk for failing to perform the actions within the required time frame is low has been added to the safe and stable discussion in the LAR Transition Report Section 4.2.1.2. Additionally, the text has been revised to indicate that the TDAFW flow can be locally throttled, that the steps to perform this action are proceduralized, and the action has been evaluated and demonstrated to be feasible. The revised LAR Transition Report Section 4.2.1.2 is provided in Attachment 1 to this Enclosure.

Page 44 of 96 to ULNRC-05851 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.

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. Attachment 1 to Callaway Plant FPRA Calculation 17671-015, NFPA 805 Fire PRA, PRA Quality Summary, provides the categorization, i.e., A, B, or C, of each of the upgrade items associated with the internal events PRA. Items that are A or B have been incorporated into the Fire PRA after they were updated for the internal events PRA. Those items categorized as C, were determined to not impact the Fire PRA used for the transition to NFPA-805.

Justification for this determination is provided in Attachment 4 to this same calculation.

b. The focused-scope internal events PRA peer review occurred in August 2011. The peer review assessed the following aspects of the Callaway Plant internal events PRA against relevant Supporting Requirements of the Standard:

i. Service Water and Component Cooling Water support system initiator fault trees, ii. Internal Flooding Analysis, iii. Interfacing Systems LOCA, Page 45 of 96 to ULNRC-05851 iv. Common Cause Failure analysis.

v. LERF modeling.

vi. Implementation of the WOG 2000 RCP Seal LOCA model.

The F&O's received from this peer review were received after the Callaway Plant LAR to adopt NFPA 805 was submitted. The table following this RAI response dispositions each F&O with respect to the IE-PRA and the FPRA. The current FPRA model has not been updated to incorporate any findings from the August 2011 Internal Events Focused Peer Review.

c. This RAI response will be provided with supplemental correspondence.

Page 46 of 96 to ULNRC-05851 PRA RAI 1 Table 1 - PWROG Focused-Scope Internal Events Peer Review F&Os and Their Disposition F&O Associated (F)inding or Brief Description Disposition for IE PRA Disposition for FPRA No. SR(s) (S)uggestion 1-1 IC-C5 F Convert SSIE and A plant availability factor of Fire PRA calculated its own IEFs.

ISLOCA IE frequencies 0.9, used for the other [n/a]

to Rx-yr basis. internal events initiators, was applied to the SSIE and ISLOCA frequency quantification.

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

justification for the use of a [n/a]

single alignment.

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

contain only one initiator/frequency BE.

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

probabilities is now used.

Therefore, this F&O has been addressed.

Page 47 of 96 to ULNRC-05851 PRA RAI 1 Table 1 - PWROG Focused-Scope Internal Events Peer Review F&Os and Their Disposition (cont'd)

F&O Associated (F)inding or Brief Description Disposition for IE PRA Disposition for FPRA No. SR(s) (S)uggestion 1-8 IE-D1, AS- F Documentation builds The three examples cited in Not applicable to the FPRA.

C1, LE-G1, on earlier the F&O were addressed.

IFSN-A12, documentation.

IFSN-B1 1-9 IE-A5 F Is loss of ESW a Further evaluation indicates Finding is not applicable to the separate IE? that creation of a separate FPRA. Loss of ESW/SW is Loss of ESW initiating event considered in fault trees for is not necessary or justified. FPRA.

1-141 DA-D3, F Treatment of CCF F&O was evaluated. F&O IE-PRA response is applicable to DA-E2 uncertainty. response includes 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 IFSO-B2, generic data for human- pending. This F&O is related the F&O are not used in the IFEV-A7 induced flooding or to the internal events IF FPRA. [n/a]

develop plant-specific analysis, and does not impact human-induced flood the Fire PRA.

frequencies.

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

IE PRA.

Page 48 of 96 to ULNRC-05851 PRA RAI 1 Table 1 - PWROG Focused-Scope Internal Events Peer Review F&Os and Their Disposition (cont'd)

F&O Associated (F)inding or Brief Description Disposition for IE PRA Disposition for FPRA No. SR(s) (S)uggestion 2-6 LE-B1 F Probability of successful MAAP runs were performed The FPRA Level 2 evaluation ex-vessel cooling should to justify the probability used. uses probabilities and split be justified. fractions from the previous Level 2 analysis. [n/a]

3-1 AS-A1, F The F&O questions The need to question seal S3 tree for FPRA is not the same AS-A2, whether the S(3) event cooling following a very event tree as for the internal AS-A3 tree should question loss small LOCA was evaluated. events. S3 event tree for FPRA of RCP seal cooling. It was determined that loss of delineates very small LOCAs seal cooling does not need to caused by fire induced events.

be included in the S(3) event Loss of seal cooling caused by tree. fire 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 related 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 Residual LOSP and consequential AS-B2 consequential LOOP in a result, consequential LOOP LOSP are not included in FPRA.

RCP seal LOCA AS was added to the Tc and Tsw (A LOSP resulting from a fire analysis. event trees. and/or random failures inside the plant boundary is included in the FPRA.) The assumption was reviewed and approved by the fire peer review. [n/a]

Page 49 of 96 to ULNRC-05851 PRA RAI 1 Table 1 - PWROG Focused-Scope Internal Events Peer Review F&Os and Their Disposition (cont'd)

F&O Associated (F)inding or Brief Description Disposition for IE PRA Disposition for FPRA No. SR(s) (S)uggestion 1-31 IE-C8 S Include PEG01A FTS Calculation EG-19 Revision CCF failures on LOSP are on loss/recovery of 1 was modified such that included in the FPRA. A power to the pump. pump start failures (including secondary fail-to-start failure common cause start failures) mode is not included for the 50%

for PEG01A (the running of the time pump A is 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 This issue pertains only to the mission time used for resolved. Loss of All CCW initiator fault CCW pressure tree/quantification. It does not transmitters in the CCW impact the Fire PRA.

initiator model/FT.

1-6 IE-C14 S Justify the 1" ISLOCA ZZ-105, Rev. 0, Add. 1 was Same justification as for IE-PRA screening criterion. revised to add justification for 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 updated probabilities identified with addressed. developed by this resolution have application of the not been implemented in the common-cause data. FPRA. Will be implemented in next FPRA update.

Page 50 of 96 to ULNRC-05851 PRA RAI 1 Table 1 - PWROG Focused-Scope Internal Events Peer Review F&Os and Their Disposition (cont'd)

F&O Associated (F)inding or Brief Description Disposition for IE PRA Disposition for FPRA No. SR(s) (S)uggestion 1-151 LE-C6, LE- S The F&O suggests This F&O has not yet been This F&O is a suggestion, which C7 consideration of pre- addressed. However, the is not yet addressed by the initiator CTMT isolation LERF analysis already internal events PRA.

failures in the CTMT includes a FAIL_LEAK Consequently, it has not been isolation systems model. event, obviating the need to addressed by the Fire PRA.

take any action in response to this F&O.

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 addressed in the FPRA.

humidity, temperature, the Internal Flooding etc., in IF analysis. Notebook.

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

enhancement. to 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 addressed in the FPRA.

enhancement relative to this 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 specific experience that F&O is related to the IE IF could impact flood analysis, and has no bearing on likelihood. the Fire PRA.

Page 51 of 96 to ULNRC-05851 PRA RAI 1 Table 1 - PWROG Focused-Scope Internal Events Peer Review F&Os and Their Disposition (cont'd)

F&O Associated (F)inding or Brief Description Disposition for IE PRA Disposition for FPRA No. SR(s) (S)uggestion 2-1 LE-A3 S Suggested LERF Additional information added IE-LERF documentation is not documentation to LERF Notebook. applicable to FPRA.

enhancement.

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

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

IFPP. F&O.

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

required for the to the LERF Notebook.

definition used for early release.

2-5 LE-C5 S Suggestion relative to Specific suggestions of the F&O not applicable to FPRA the use of conservative F&O were addressed. LERF analysis.

versus 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 IE-PRA event tree documentation tables to match current the AS calculation set to not applicable to FPRA.

Tc and Tsw event trees. address this F&O.

Page 52 of 96 to ULNRC-05851 PRA RAI 1 Table 1 - PWROG Focused-Scope Internal Events Peer Review F&Os and Their Disposition (cont'd)

F&O Associated (F)inding or Brief Description Disposition for IE PRA Disposition for FPRA No. SR(s) (S)uggestion 3-5 AS-A7 S Suggestion to consider Suggestion was evaluated, Spurious-open PORV is a the need to add the and justification for not consequential event in the FPRA, potential for a stuck- adding a stuck open PORV to which will appear in any scenario open PORV to T1s, Tc these ETs was generated. where fire damage can cause it to and Tsw event trees. occur. [n/a]

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

documentation. pertains 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 Random flooding events are not IFSN-A8 floor penetrations and been addressed. However, it addressed in the FPRA.

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

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

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

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

text. this F&O.

4-2 LE-C8 S Suggestion for Information added to LERF F&O is not applicable to the additional discussion of Notebook. FPRA.

how Level 1 and Level 2 models are linked.

Page 53 of 96 to ULNRC-05851 PRA RAI 1 Table 1 - PWROG Focused-Scope Internal Events Peer Review F&Os and Their Disposition (cont'd)

F&O Associated (F)inding or Brief Description Disposition for IE PRA Disposition for FPRA No. SR(s) (S)uggestion 4-3 AS-A7, S Suggested enhancement Additional text justification The event tree assumptions that AS-A10 of RCP seal LOCA was developed, and will be are critiqued here are not made in accident sequence included in an AS Notebook, the FPRA. F&O does not apply documentation. currently under development. to the FPRA.

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

enhancement.

4-5 LE-C9 S Suggestion to justify Justification/text added to the F&O not applicable to the FPRA.

feedwater availability LERF Notebook.

after core damage (as credited in the LERF model).

Note 1 - These F&Os will be evaluated and/or incorporated into the fire PRA during the next update, per LAR Transition Report Attachment S Implementation Item 12-805-001. The revised LAR Transition Report Attachment S is provided as Attachment S to this Enclosure.

Page 54 of 96 to ULNRC-05851 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 Probabilistic Risk Assessment RAI 02 Fire PRA Peer Review:

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

Internal Event Peer Review:

The internal events PRA peer review occurred in May 2006. The peer review used the ASME-RA-Sb-2005 (Dec. 2005) version of the PRA standard. This version of the standard incorporated NRC comments from RG 1.200, Trial for Use, Attachment A (Jan. 2004). As such, the internal events PRA was peer reviewed against the clarifications and qualifications presented in the latest revision of RG 1.200 available at the time of the review. 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. This is being tracked by Implementation Item 12-805-001 as shown in Attachment S to this Enclosure.

Page 55 of 96 to ULNRC-05851 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.

Response to Probabilistic Risk Assessment RAI-03 This RAI response will be provided with supplemental correspondence.

Page 56 of 96 to ULNRC-05851 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.)

Page 57 of 96 to ULNRC-05851 Response to Probabilistic Risk Assessment RAI 04

a. The current status of CCF modeling in the internal events PRA with respect to the issues in the F&O is summarized below. This status also applies to the Fire PRA, which used the internal events fault trees as a basis.

1) Breakers for pumps and diesel generators are considered to be within the component boundary of the component. Component boundary definitions from NUREG/CR-6928 were used. A specific CCF term for these breakers is not necessary.

2) Bus load breakers for offsite power distribution do not have a common cause event, except for the two breakers which supply alternate power to the 4160V emergency busses. These breakers are modeled with a CCF event.

3) Battery chargers are modeled with CCF.

4) The recent PRA update for the internal events used CCF equations from NUREG/CR-5485. The CCF factors and values were derived from WCAP-16607-NP, Rev 1.

b. The internal events PRA has been updated to address these event probabilities. This update calculated new probabilities for the events in question using methods that meet the requirements of SR DA-D2. Specifically, plant-specific strainer failure data was gathered and used to Bayesian-update appropriate generic data sources. The failure mode for strainer fail to run was eliminated and the failure mode for strainer fail to start was changed to incorporate all failures. Additionally, a calculation was developed to determine an updated probability for the HYDRAULICSYSFAIL special event. The internal events PRA is complete and meets the requirements of SR-DA-D2.

The Fire PRA uses the updated data for STR-FS and HYRAULICSYSFAIL. The basic event for STR-FR has been retained in the Fire PRA at its original value. However, the probability of STR-FR is low enough that the event does not appear in the dominant sequences. Therefore, the Fire PRA is consistent with the internal events PRA.

F&O DA-3 has now been closed for both the internal events PRA and the Fire PRA.

c. This F&O is not applicable to the Fire PRA. The Fire PRA does not credit any recovery terms used in the internal events PRA. The recovery terms in the internal events-PRA represent restoration of a system to operability through maintenance, replacement and repair.

These actions are not modeled in the Fire PRA.

Fire PRA models NFPA 805 recovery actions which involve changing the position of a valve or a breaker. In all cases, these recovery actions are clearly identified and evaluated for the risk of taking this action.

Page 58 of 96 to ULNRC-05851

d. The Fire PRA performed a detailed, fire-specific review of potential fire-induced ISLOCA pathways. An explanation of the process is outlined in Section 3.5 of the Component Selection Report (17671-002) and Section 4.6 of the Fire Induced Risk Model Report (17671-004). Fire-specific screening criteria were developed, all internal events ISLOCA paths were re-considered, and all containment penetrations with high-low pressure interfaces were evaluated.

The Fire PRA Peer Review in October 2009 reviewed the fire-induced ISLOCA model.

Specific references to the fire-induced ISLOCA analysis are presented in the summary of assessment for supporting requirements ES-A6 and PRM-B14 in the final Peer Review Report (LTR-RAM-II-10-019, Rev. 0). The discussion associated with these two SRs indicates that the Peer Review team acknowledged the fire-specific ISLOCA model and had reviewed the associated documentation.

In addition, the fire PRA peer review team reviewed the MSO report, which contains ISLOCA modeling documentation in the resolution of MSO issues 10, 14, and 15. This report documents the potential ISLOCA pathways and how they were treated in the Callaway Plant fire-induced ISLOCA model. There are no F&Os directly related to the fire-induced ISLOCA modeling.

e. Fire induced floods can occur due to three mechanistic scenarios within the scope of the Fire PRA. Occurrence of a random pipe break concurrent with a fire is not considered in the Fire PRA. In all instances where there was a potential for fire induced flood, the potential for flooding damage was evaluated and considered to be risk-insignificant compared to the damage caused by the fire. The three mechanistic scenarios are:

1) Fire sprinkler actuation. The potential for flooding due to fire sprinkler actuation is evaluated as part of the fire protection design basis. Scenarios for sprinkler flooding are not explicitly included in the Fire PRA.

2) Overfill of a subcooled water system, causing opening of a relief valve in that system.

There is one scenario for potential flooding through a relief valve discharging into the plant. This is from the relief valves on the Component Cooling Water (CCW) surge tanks. Each surge tank is provided with a relief valve and a vacuum breaker. The Fire PRA contains scenarios for CCW overfill and water flow through the relief valve on the surge tank. These scenarios already fail the affected CCW train. A recovery action (OP-OMA-FF-ISOEG) is credited to isolate the flow path and prevent failure of the unaffected train. If the action is successful, the failure is confined to the affected train.

Risk of failure of the action is evaluated as part of the fire risk evaluation process.

3) Fire causing failure of non-metallic pipe. There were two instances of this failure mechanism, which will be discussed in the response to Probabilistic Rick Assessment RAI 4(f).

f. This RAI response will be provided with supplemental correspondence.

Page 59 of 96 to ULNRC-05851 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 The Fire PRA was peer reviewed in October 2009. The only changes to the Fire PRA model between the peer review and the NFPA 805 LAR submittal were to address Fire PRA peer review comments, to expand the system fault tree and plant event tree models to address plant modifications documented in Attachment S of the Transition Report, and to refine the modeling in individual areas based on feedback from the FSA/FRE process. There were no new methods or modeling approaches, as defined in Chapter 1 of the combined PRA Standard, that have been implemented in the Callaway Plant Fire PRA since the October 2009 peer review. Thus, there has not been a need for a focused scope peer review of the Fire PRA.

Many of the changes to address peer review comments were to change the values for failure data, such as changes to spurious actuation probabilities. However, some changes were made to the model structure to reflect plant modifications. The plant configuration at the time of the Fire PRA peer review did not include the Non-Safety Auxiliary Feedwater Pump (AFW) pump nor the electrical power connection from the Alternate Emergency Power System (AEPS) diesel generators via the offsite connection. These changes were incorporated into the Fire PRA using the same methods and modeling approaches that had been examined in the October 2009 peer review. These involved fault tree /event tree changes to the existing PRA models to incorporate plant changes. Additionally, the human reliability analysis was updated to address changes to plant procedures. All of these Fire PRA changes are clearly within the definition of PRA Maintenance as defined in Section 1-2.2 of the combined PRA standard and do not require a focused scope peer review.

Page 60 of 96 to ULNRC-05851 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 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

a. F&O ES-A1-1. As the Fire PRA was finalized for the NFPA 805 LAR submittal, the MSO list used in the Fire PRA was updated using WCAP-16933-NP Rev 1, PWR generic List of Fire induced Multiple Spurious Operation Scenarios, dated June 2010.

b. F&O ES-B1-1 was a SUGGESTION level F&O. The response to this F&O was to clarify the references. The Fire PRA Peer Review team observation was that the Task 2 report did not Page 61 of 96 to ULNRC-05851 include a list of safe shutdown equipment list (SSEL) components that are included in the Fire PRA. The observation continues to say this list was provided in the Task 4 report (Fire-Induced Risk Model Development). The suggestion was that the list of SSEL components included in the Fire PRA should be included in the Task 2 report, so that the Task 3 work can commence.

The Fire PRA tasks are not done in isolation and it makes no difference where this list is documented, as long as it is properly referenced. Additionally, the task 3 cable selection was for all SSEL components, regardless if they are included in the PRA.

c. F&O ES-B2-1. This was a SUGGESTION level F&O. The Callaway Plant MSO search for flow diversion paths did not rely on the internal events PRA. The MSO Expert panel had been supplied with instructions and methods for identifying MSO scenarios before the MSO Expert Panel meeting. These were documented in Attachments A through C of the updated MSO Expert Panel report, Callaway-FPRA-17671-002b, MSO Expert Panel, used in the Fire PRA update to support the Callaway Plant NFPA 805 LAR. Specifically, Attachment C of the updated MSO Expert Panel report provides instructions on the search for diversion paths.

These instructions were not included as part of the MSO Expert Panel report for the Fire PRA Peer Review.

d. F&O ES-C1-1. ES-C1-1 was a SUGGESTION level F&O. ES-C1-1 involves the documentation of which instruments are required to support each HFE. Instrumentation for operator actions was traced but was not clearly documented in the Component Selection report reviewed by the Fire PRA Peer Review team. Following the Fire PRA Peer Review, the instrumentation traced for each HFE was provided in Table 3-3 of the Fire PRA HRA Report, Callaway-FPRA-17671-011- Fire HRA.

e. F&O CS-B1-1. CS-B1-1 is a FINDING level F&O. This F&O was dispositioned by conducting additional analysis. CS-B1-1 involves assurance that all electrical busses required for operation of the PRA components have been identified for over current coordination and protection evaluation. The response to the F&O follows:

The results of the associated circuits review for NFPA 805 NSCA, NPO and the Fire PRA identified that additional breaker/fuse coordination activities would be needed to address electrical power alignments associated with offsite power (that was being credited for Safe Shutdown, Non-power Operations, and Fire PRA), and to improve plant documentation and controls with respect to maintaining the engineering basis for series fuse coordination. A new calculation, ZZ-547, "NFPA 805 Breaker/Fuse Coordination Study for PB, PG, PK, and PN Circuits for Safe Shutdown Equipment",

was issued. Calculation ZZ-547 was reviewed and included with the other previously existing breaker/fuse calculations for Callaway Plant as part of the associated circuits review for NFPA 805 NSCA, NPO and the Fire PRA. The results of this review are summarized in Section 8.7 of Calculation KC-26, "Nuclear Safety Capability Assessment". The results of the review did not identify any model or analysis changes due to lack of breaker/fuse coordination.

Page 62 of 96 to ULNRC-05851

f. FSS-B01-2 is a FINDING level F&O. FSS-B01-2 indicates that at the time of the peer review, the control room analysis for cabinet fires did not consider that a) fires in the HVAC control cabinet would fail HVAC, changing the conditional probability of forced evacuation, b) cabinet fires in the solid state protection system (SSPS) cabinets could lead to a spurious safety injection actuation or loss of offsite power, which could complicate the plant response and c) cabinet fires could spread to secondary combustibles.

In response to part (a) of this F&O, the control room fire analysis was updated to consider that fires originating in the HVAC control cabinet would fail HVAC and result in a higher evacuation probability than if the HVAC was available.

In response to part (b) of this F&O, all cabinets in the Equipment Cabinet Area (ECA) were assessed for their capability to cause the following:

a) Spurious safety injection actuation, b) Loss of offsite power, c) Loss of control room HVAC.

In addition, a CFAST run was developed in response to part (c) to model fire in a single cabinet, spreading to two adjacent cabinets on either side in 10 minutes.

The scenario responses for these cabinet fire groups were modified to include the effect of the failed equipment in the cabinet. This process is documented in Callaway-FPRA-17671-010b, Main Control Room Fire Analysis.

Page 63 of 96 to ULNRC-05851 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.

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.

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 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.

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.

Page 64 of 96 to ULNRC-05851 Response to Probabilistic Risk Assessment RAI 07

a. This RAI response will be provided with supplemental correspondence..

b. NUREG/CR-6850 suggests that the use of a single overall failure probability value to represent the failure of reaching safe shutdown using alternate means can be used if the probability value is evaluated conservatively and a proper basis is provided. It notes that this approach was used in several IPEEE submittals and that in many cases, 0.1 was used as a point value estimate for the probability. Additionally, section 5.1.3 of NUREG-1921 (March 2012 Draft for Full Committee ACRS meeting) further supports the use of 0.1 as follows:

NUREG/CR-6850 [1] suggests that the use of a single overall failure probability value to represent the failure of reaching safe shutdown using alternative means can be used if the probability value is evaluated conservatively and a proper basis is provided. It notes that this approach was used in several IPEEE submittals and that, in many cases, 0.1 was used as a point value estimate for the probability. Before crediting this approach, the analyst must have applied the criteria discussed in Section 4.3 for assessing the feasibility of the operator action(s) associated with that HFE.

Section 4.3 of NUREG-1921 describes the feasibility assessment and consideration of the following items, which the Callaway Plant satisfies. NUREG-1921 also recognizes the estimation of Main Control Room (MCR) evacuation as an area for future research.

• Required actions are proceduralized or skill-of-the-craft

• Sufficient time is available to perform the required actions

• Sufficient manpower is available

• Operators are trained on the required actions

• Required tools and parts are available

• Areas where actions are required are accessible

• Necessary and sufficient cues and indications are available The Callaway Plant Nuclear Safety Capability Assessment (calculation KC-26) lists all of the recovery actions taken when implementing the Main Control Room inaccessibility procedure.

This same calculation also documents the feasibility of these actions within the context of the seven bulleted items from NUREG-1921, listed above. While there are potentially many actions to be taken to support achieving a safe and stable condition, these local actions (defined as actions where the operators leave the main control room) are not intrinsically different from other local actions that may be performed following other accident initiators, such as Station Black Out (SBO). Detailed human error probabilities (HEP) in SBO scenarios are typically of the order 1E-02, and an HEP associated with a detailed evaluation of individual, local actions following MCR evacuation would be of the same order. In addition to the local, manual portion of the HEP (estimated to be ~5E-2), the 0.1 MCR evacuation HEP includes failure to diagnose the need for MCR evacuation, and failure to conduct the evacuation in a timely Page 65 of 96 to ULNRC-05851 manner. For scenarios that required MCRevacuation, an HEP of 0.1 was assigned as the total probability of failing to evacuate and establish local control successfully.

c. This RAI response will be provided with supplemental correspondence.

d. The Callaway Plant Fire HRA considers dependencies between human failure events (HFEs) in two ways. First, a review of the operator actions was conducted during the model development and those HFEs that were functionally similar were identified, and a separate HFE was added to the Fire PRA to explicitly capture dependencies. The explicitly modeled dependent event is termed the common cognitive HFE as its probability is primarily driven by a failure to recognize the plant condition or failure to recognize the proper procedure in order to respond to the plant condition. The complementary portion of the operator actions (termed execution HFE) are, by definition, considered independent and have been included explicitly in the Callaway Plant Fire PRA as separate basic events.

Second, even though the execution HFEs are modeled as separate basic events, the human error probability (HEP) development for the execution HFEs takes into account preceding operator actions. Specifically, the HEP for the execution HFEs has been calculated using a timeline that takes into account all preceding operator actions in the Fire Response Procedure (whether these operator actions are in the Fire PRA model or not). The timing analysis, and resulting HEP for each execution HFE, accounts for the other execution HFEs in the same scenario. Thus, if an early operator action takes a long time to perform then the failure probability for a later HFE would typically be increased if the time margin is reduced.

This peer review observation was a SUGGESTION level F&O to conduct a review and ensure that combinations of execution HFEs that occur in the Fire PRA 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 used in the LAR accounts for dependencies in these 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 Attachment S, Table S-3 in the Callaway Plant NFPA 805 License Amendment Request.

e. This RAI response will be provided with supplemental correspondence.

f. This RAI response will be provided with supplemental correspondence.

Page 66 of 96 to ULNRC-05851 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.

Response to Probabilistic Risk Assessment RAI 08

a. Fire Influencing Factors (FIFs) of less than 1.0 were used for the Storage and Maintenance terms in fire areas that are administratively controlled as No Storage and No Hot-work areas while the plant is in Modes 1 and 2. However, in these areas the Occupancy term was always held to at least a value of 1.0, such that the total transient area weighting factor is always greater than 1.0.

Since the minimum combined weight of all locales with special weighting factors is always greater than 1.0, there is no need for a sensitivity study that addresses special weighting factor locales with a combined weight of less than 1.0.

The minimum FIFs used at Callaway Plant are representative of a transient combustible free zone with hot-work prohibited while in Modes 1 and 2. The hot-work FIF is assigned a minimum value Page 67 of 96 to ULNRC-05851 of 0.05 to account for procedural violations, and the storage FIF was assigned a minimum value of 0.1 to account for procedural violations. Thus the minimum FIF summation was 1.15.

b. This RAI response will be provided with supplemental correspondence.

Page 68 of 96 to ULNRC-05851 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.

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.

Response to Probabilistic Risk Assessment RAI 09 This RAI response will be provided with supplemental correspondence.

Page 69 of 96 to ULNRC-05851 Probabilistic Risk Assessment RAI 10 FAQ 08-0052 (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).

Response to Probabilistic Risk Assessment RAI 10 This RAI response will be provided with supplemental correspondence.

Page 70 of 96 to ULNRC-05851 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 The bulleted items on the list are all controls/indications provided on the Auxiliary Shutdown Panel (RP118B). There are no other primary control stations defined for Callaway Plant. The bulleted list in LAR Transition Report Attachment G has been modified to more clearly reflect this. The revised portion of LAR Transition Report Attachment G is provided as Attachment G to this Enclosure.

Page 71 of 96 to ULNRC-05851 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).

Response to Probabilistic Risk Assessment RAI 12 This RAI response will be provided with supplemental correspondence.

Page 72 of 96 to ULNRC-05851 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.

Response to Probabilistic Risk Assessment RAI 13 This RAI response will be provided with supplemental correspondence.

Page 73 of 96 to ULNRC-05851 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 In the LAR Transition Report Attachment W, Section W.2 there is an incorrect sentence regarding self approval. Ameren Missouri understands NRC approval is necessary before the transition to NFPA 805 can be completed. The referenced sentence has been deleted from Attachment W, Section W.2 as provided in Attachment W to this Enclosure.

Page 74 of 96 to ULNRC-05851 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 The title of Attachment W, Table W-1 in the LAR Transition Report is incorrect. The introduction of the table in LAR Attachment W, Section W.1 provides the correct description of the table contents.

The title should read:

Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF)

The Callaway Plant Fire PRA model contains over 1200 individual fire scenarios. Of these, 197 scenarios account for the top 95% of CDF. Many of the scenarios in the cumulative 95% have low individual values. Thus, the criteria used for developing Attachment W, Table W-1 was to include only those scenarios contributing to greater than 1% of CDF per the definition of significant accident sequences in the combined PRA standard. The revised title for Attachment W, Table W-1 is shown in Attachment W of this Enclosure.

Page 75 of 96 to ULNRC-05851 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.

Response to Probabilistic Risk Assessment RAI 16 This RAI response will be provided with supplemental correspondence.

Page 76 of 96 to ULNRC-05851 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.

Response to Probabilistic Risk Assessment RAI 17 This RAI response will be provided with supplemental correspondence.

Page 77 of 96 to ULNRC-05851 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 The methodology that was used at Callaway Plant to evaluate defense-in-depth is documented in the NFPA 805 Task Plan for Fire Risk Evaluations, Revision 1, dated April 2011. The methodology is as follows.

Defense in Depth Approach A review of the impact of the VFDRs on defense-in-depth was performed, regardless of the risk evaluation method used. The review of defense-in-depth is typically qualitative and should address each of the elements with respect to the proposed change.

1) Evaluate the fire area for the impact of the VFDRs on fire protection defense-in-depth. Fire protection defense-in-depth is achieved when an adequate balance of each of the following elements is provided:

a. Preventing fires from starting;

b. Rapidly detecting fires and controlling and extinguishing promptly those fires that do occur, thereby limiting fire damage; and

c. Providing an adequate level of fire protection for structures, systems, and components important to safety, so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed.

2) In general, the defense-in-depth requirement is satisfied if the proposed change does not result in a substantial imbalance among these elements. Table 1 that follows contains additional defense-in-depth guidance.

3) In evaluating defense-in-depth, it may become necessary to consider the potential for risk significant fire scenarios to impact VFDRs. A fire scenario is defined as a unique quantification of a fire damage state (which may include severity factors and probability of non-suppression) multiplied by a CCDP or CLERP to arrive at a CDF or LERF. For purposes of defense-in-depth, potentially risk significant fire scenarios could be characterized as follows, for example:

Page 78 of 96 to ULNRC-05851 x A scenario in which the calculated risk is equal to or greater than 1E-6/year for CDF and/or 1E-7/year for LERF, could be characterized as potentially risk significant.

x A scenario in which the calculated risk falls between 1E-6/year and 1E-8/year for CDF, or between 1E-7/year and 1E-9/year for LERF, and where DID echelon 1 and 2 attributes are causing a significant reduction in risk, could be characterized as potentially risk significant .

x A scenario in which the calculated risk is less than 1E-8/year for CDF and/or 1E-09 for LERF, regardless of reliance on DID echelon 1 and 2 attributes, may be characterized as potentially not risk significant. These values are considered potentially not risk significant based on being two to three orders of magnitude below the acceptance criteria of RG 1.174 as referenced by RG 1.205, Rev. 1.

x A scenario with a high consequence (i.e., CCDP>E-1) could be considered potentially risk significant.

4) Fire protection features and systems relied upon to ensure defense-in-depth should be clearly identified in the assessment (e.g., detection, suppression system, etc.).

5) Verify that defense-in-depth is maintained by assessing and documenting that the balance is preserved among prevention of core damage, prevention of containment failure, and mitigation of consequences. Regulatory Guide 1.174 provides guidance on maintaining the philosophy of nuclear safety defense-in-depth that is acceptable for NFPA 805 Fire Risk Evaluations.

6) Each fire area shall be evaluated for the need to incorporate defense-in-depth enhancements to provide assurance that plant performance goals can be achieved and maintained.

Documentation of these defense-in-depth enhancements can be on a fire area basis and/or tied directly to a VFDR disposition, as appropriate.

7) Provide the results of the defense-in-depth review in a tabular format, such as that shown in the example in Table 2 that follows. Defense-in-depth attributes shall be evaluated for applicability to NFPA 805, Section 4.2.3 or 4.2.4 (Ch. 3, as required).

x If a defense-in-depth attribute is credited for NSCA deterministic criteria, licensing action or engineering equivalency evaluation then the system/feature should already be considered to form an integral part of defense-in-depth. The parent echelon of the system/feature should then be evaluated against the process and considerations presented in Table 1 that follows, to determine if any improvements or changes are necessary, such as to offset a weakness in another echelon.

x If the Fire PRA credits any of the fire protection features or a recovery action to improve the risk profile then these attributes or features should already be considered to form an integral part of defense-in-depth. The parent echelon of the system/feature should then be evaluated against the process and considerations presented in Table 1 that follows, to Page 79 of 96 to ULNRC-05851 determine if any improvements or changes are necessary, such as to offset a weakness in another echelon.

x Defense-in-depth attributes that go above and beyond the existing requirement(s) with the purpose of bolstering derived weaknesses within the defense-in-depth elements to maintain an overall balance should be designated as a change or improvement necessary for defense-in-depth.

o Note - this may or may not involve a physical improvement to the element, but by virtue of including an attribute that was not required for deterministic or risk reasons, defense-in-depth is considered enhanced.

x Features or enhancements required for defense-in-depth warrant consideration for inclusion in the monitoring program.

Page 80 of 96

Enclosure 1 to ULNRC-05851 Table 1 - Considerations for Defense-in-Depth Determination Method of Providing DID Considerations Echelon 1: Prevent fires from starting Combustible Material Combustible and hot work controls are fundamental elements of defense-in- depth and as such are always in place. The issue to be considered Controls during the fire risk evaluation is whether this element needs to be strengthened to offset a weakness in another echelon thereby providing a Hot Work Controls reasonable balance.

Considerations include:

Creating a new Transient Combustible Free Area Creating a new Hot Work Restriction Area Modifying an existing Transient Combustible Free Area or Hot Work Restriction Area The fire scenarios involved in the fire risk evaluation quantitative calculation should be reviewed to determine if additional controls should be added.

Review the remaining elements of defense-in-depth to ensure an over-reliance is not placed on programmatic activities for weaknesses in plant design.

Page 81 of 96

Enclosure 1 to ULNRC-05851 Table 1 - Considerations for Defense-in-Depth Determination Echelon 2: Rapidly detect, control and extinguish promptly those fires that do occur thereby limiting fire damage Detection system Automatic suppression and detection may or may not exist in the fire area of concern. The issue to be considered during the fire risk Automatic fire suppression evaluation is whether installed suppression and or detection is required for defense-in-depth or whether suppression/detection needs to be Portable fire extinguishers strengthened to offset a weakness in another echelon thereby providing a reasonable balance.

provided for the area Hose stations and hydrants Considerations include:

provided for the area Fire Pre-Fire Plan Risk Insights:

If the variance is never affected in a potentially risk significant fire scenario, manual suppression capability may be adequate and no additional systems required.

Recovery Actions:

If the fire area requires recovery actions, then as a minimum, detection and manual suppression capability are required, and suppression should be considered.

If a fire area contains neither suppression nor detection and a recovery action is required, consider adding detection and/or suppression.

Firefighting Activities:

If firefighting activities in the fire area are expected to be challenging (either due to the nature of the fire scenario or accessibility to the fire location) then both suppression and detection may be required Fire Scenarios:

If fire scenarios credit fire detection or fire suppression systems, then these should be considered to form an integral part of defense-in-depth Page 82 of 96

Enclosure 1 to ULNRC-05851 Table 1 - Considerations for Defense-in-Depth Determination Echelon 3: Provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed Walls, floors ceilings and structural elements are rated If fires occur and they are not rapidly detected and promptly extinguished, then the third echelon of defense-in-or have been evaluated as adequate for the hazard. depth would be relied upon.

Penetrations in the fire area barrier are rated or have The issue to be considered during the fire risk evaluation is whether existing separation is adequate (or over relied been evaluated as adequate for the hazard. on) and whether additional measures (e.g., supplemental barriers, fire rated cable, or recovery actions) are Supplemental barriers (e.g., ERFBS, cable tray covers, required to offset a weakness in another echelon thereby providing a reasonable balance.

combustible liquid dikes/drains, etc.) Considerations include:

Fire rated cable Guidance provided to operations personnel detailing Risk Insights:

the required success path(s) including recovery If the variance is never affected in a potentially risk significant fire scenario, internal fire area actions to achieve nuclear safety performance criteria. separation may be adequate and no additional reliance on recovery actions necessary.

If the variance is affected in a risk significant fire scenario, internal fire area separation may not be adequate and reliance on a recovery action, supplemental barrier, or other modification may be necessary.

If the consequence associated with the variance is considered high (e.g., CCDP>1E-01 or by qualitative SSD assessment) regardless of whether it is in a risk significant fire scenario, a recovery action, supplemental barriers, or other modification should be considered.

There are known modeling differences between a Fire PRA and nuclear safety capability assessment due to different success criteria, end states, etc. Although a variance may be associated with a function that is not considered a significant contribution to core damage frequency, the variance may be considered important enough to the NSCA to retain as a recovery action.1 Operations Insights:

If the sequence to perform a recovery action is particularly challenging then including the action for defense-in-depth may be considered.2 The fire scenarios involved in the fire risk evaluation quantitative calculation should be reviewed to determine the fires evaluated and the consequence in the area to best determine options for this element of defense in depth.

1 An example would be components in the NSCA associated with maintaining natural circulation at a pressurized water reactor that are not modeled explicitly in the Fire PRA since they are not part of a core damage sequence.

2 An example would be a recovery action that is unique in nature, time critical and/or not included in emergency response procedures such that the MCR staff may not be able to quickly recognize and perform the required action.

Page 83 of 96

Enclosure 1 to ULNRC-05851 Table 2: Example Defense-in-Depth Impact Review for Fire Area Required to Support Changes or Deterministic Improvements Method of Providing DID Basis/Justification Analysis or Necessary for Fire DID?

PRA?

Echelon 1: Prevent fires from starting Combustible Control is implemented in accordance with Procedure X, "Control of Combustible Yes No Materials. x This element is adequate based on no perceived weakness of, or over-reliance on, another echelon of defense-in-depth.

Hot Work Control is implemented in accordance with Procedure X, "Welding, Burning, and Yes No Grinding Activities" Echelon 2: Rapidly detect, control and extinguish promptly those fires that do occur thereby limiting fire damage Fire Detection System No No x Detection is not credited in the performance-based analysis, firefighting Automatic Fire Suppression Yes No activities are not expected to be challenging, and no recovery actions are required; therefore, no change or improvement to the installed system is required to maintain defense-in-depth.

Portable Fire Extinguishers Yes No x Automatic suppression is credited in the performance based analysis. No Hose stations and hydrants located in the area(s) Yes No further change or improvement to the installed system is required to maintain defense-in-depth.

Fire Pre-Fire Plan Yes No Echelon 3: Provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed Walls, floors ceilings and structural elements are x Supplemental barriers are credited in the performance-based analysis, and rated or have been evaluated as adequate for the Yes No therefore, form an integral part of defense-in-depth.

hazard.

x The variance is never affected in a risk significant fire scenario and internal fire Openings in the fire area barrier are rated or have area separation is adequate. No additional Echelon 3 attributes are necessary to Yes No been evaluated as adequate for the hazard. maintain defense-in-depth.

Page 84 of 96

Enclosure 1 to ULNRC-05851 Table 2: Example Defense-in-Depth Impact Review for Fire Area Required to Support Changes or Deterministic Improvements Method of Providing DID Basis/Justification Analysis or Necessary for Fire DID?

PRA?

x There are no significant modeling differences between the Fire PRA and nuclear Supplemental barriers (e.g., ERFBS, cable tray Yes No safety capability assessment (i.e., due to different success criteria, end states, covers, etc.)

etc.) that are contributing to reduce core damage frequency.

Fire rated cable No No Guidance provided to operations personnel detailing the required success path(s) including No No recovery actions to achieve nuclear safety performance criteria.

Page 85 of 96 to ULNRC-05851 Safety Margin Approach The methodology that was used to evaluate safety margins is described the NFPA 805 Task Plan for Fire Risk Evaluations, Revision 1, dated April 2011. The methodology is summarized as follows:

Based on NEI 04-02, the requirements related to Safety Margins for the change analysis is described for each of the specific analysis types used in support of the fire risk assessment. The specific Safety Margin evaluation will depend on the change set.

The evaluation addresses whether:

x Codes and Standards or their alternatives accepted for use by the NRC are met, and x Safety analysis acceptance criteria in the licensing basis (e.g., FSAR, supporting analyses) are met, or provides sufficient margin to account for analysis and data uncertainty.

These evaluations can be grouped into categories. These categories are:

1. Fire Modeling

2. Plant System Performance

3. PRA Logic Model

4. Other

1) Fire Modeling If a performance based approach is used, the margin between the parameters describing the MEFS and the LFS and the process of judging the adequacy of that fire modeling margin is required for the overall safety margin consideration. The level of review to be performed as part of the safety margin treatment considered here involves the integration of that margin with the potential consequences of the upset, or damage, that may occur given the LFS. The acceptability of the fire modeling margin between MEFS and LFS needs to be judged in the context of the potential severity of the resulting plant system impact if an LFS were to occur. An LFS that causes an inter-system loss of coolant accident (ISLOCA) event would tend to demand a higher margin between MEFS and LFS as compared to an event that causes a degradation of long term decay heat removal.

2) Plant System Performance The development of the fire risk assessment may involve the re-examination of plant system performance given the specific demands associated with the postulated fire event. The methods, input parameters, and acceptance criteria used in these analyses needs to be reviewed against that used for the plant design basis events.

This subtask evaluates the plant system performance given the specific demands associated with the postulated fire event. The methods, input parameters, and acceptance criteria utilized in the risk-informed, performance-based analysis will be reviewed against the plant design basis events. This Page 86 of 96 to ULNRC-05851 evaluation will determine if the safety margin established in the plant design basis events is preserved.

3) PRA Logic Model This subtask evaluates results of the Fire PRA model to verify that the safety margins have not changed. The CDF and LERF importance measures of components in the cutset results will be evaluated to verify that events with high importance values have reasonable failure probabilities for the scenarios of interest. This will be particularly important for human error basic events. The results of each risk evaluation will be evaluated against the base case fire results to determine that no single event has undue influence on the results of the change analysis. This evaluation will demonstrate that the safety margin established in the PRA model is preserved and that the Fire PRA model is sufficient to treat the fire-induced core damage sequences.

4) Other (referred to as Miscellaneous in NEI 04-02)

This category addresses any other analyses not addressed above. The general requirements related to codes and standards, and acceptance criteria, provided above apply.

Example of a typical safety margin review as contained in a Fire Safety Analysis report for a Fire Area with one or more VFDRs:

In accordance with NEI 04-02, the maintenance of adequate Safety Margin is assessed by the consideration categories of analyses utilized by this Fire Risk Evaluation.

Safety margins are considered to be maintained if:

• Codes and Standards or their alternatives accepted for use by the NRC are met.

AND

• Safety analyses acceptance criteria in the licensing basis (e.g., FSAR, supporting analyses) are met, or provides sufficient margin to account for analysis and data uncertainty.

The following summarizes the bases for ensuring the maintenance of safety margins:

• The risk-informed, performance based processes utilized are based upon NFPA 805, 2001 edition, endorsed by the NRC in 10 CFR 50.48(c).

• The Fire Risk Evaluation process is in accordance with NEI 04-02, Revision 2, which is endorsed by the NRC in Regulatory Guide 1.205, Revision 1.

• The Fire PRA is developed in accordance with NUREG/CR-6850, which was developed jointly between the NRC and EPRI.

Page 87 of 96 to ULNRC-05851

• The Fire PRA has undergone an industry peer review, in order to ensure the Fire PRA meets the appropriate quality standards of ASME/ANS RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Dated February 2, 2009

• The combined analysis approach is used during transition (NEI 04-02, Section 5.3.4.3);

therefore, MEFS/LFS is not analyzed separately from the Fire PRA results.

• The Callaway Plant internal events PRA model received three peer reviews conducted in accordance with applicable NEI/ ASME standards in force at the time. The initial peer review was performed by the Westinghouse Owners Group in 2001. A full scope peer review was performed in 2006 prior to the start of the NFPA 805 transition project. This peer review used the ASME RA-Sb-2005 PRA standard. A focused scope peer review was performed by the Westinghouse Owners Group in August 2011. The 2001 and 2006 peer reviews covered all aspects of the Callaway Plant PRA model and the administrative processes used to maintain and update the model. The Callaway Plant PRA model has been revised to address all significant issues (i.e., Category A and B Facts & Observations) identified during both the 2001 and 2006 reviews.

• Fire protection systems and features determined to be required by NFPA 805 Chapter 4 have been confirmed to meet the requirements of NFPA 805 Chapter 3 and their associated referenced codes and listings, or provided with acceptable alternatives using processes accepted for use by the NRC (i.e., FAQ 06- 0008, FAQ 06-0004, 07-0033).

• Fire modeling performed 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. In general, the fire modeling performed in support of the Fire Risk Evaluations has been performed using conservative methods and input parameters that are based upon NUREG/CR-6850 as documented in the Detailed Fire Modeling Report, section 7.2. While this is generally not ideal in the context of best estimate probabilistic risk analysis, it is a pragmatic approach 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.

• In accordance with the requirements of 10 CFR 50.48(c)(iii), the Fire PRA results, including cutsets for the scenarios of concern, have been reviewed and it was verified that the results presented above do not rely solely on feed and bleed as the fire-protected safe shutdown path for maintaining reactor coolant inventory, pressure control, and decay heat removal capability for this fire area.

Evaluation of Multiple Recovery Actions This section describes the treatment of recovery actions, specifically the evaluation of feasibility and reliability of multiple, time-critical, or complex recovery actions for a particular fire scenario Page 88 of 96 to ULNRC-05851 and how the risk evaluations for recovery action probabilities consider multiple actions in a single scenario.

As required by NFPA 805 Section 4.2.4.1.6 all Recovery Actions (RAs) credited to meet the Nuclear Safety Performance Criteria must be demonstrated to be feasible. As part of determining RA feasibility a thermal-hydraulic calculation was developed to identify allowed RA completion times by fire area.

The RA feasibility evaluation then was performed (and documented in KC-26) by considering the expected plant response for a given fire area in conjunction with the following eleven (11) criteria

1) Draft Fire Procedures, 2) Systems and Indications, 3) Tools-Equipment, 4) Communications, 5)

Emergency Lighting, 6) Demonstrations, 7) Actions in the Fire Area, 8) Time, 9) Staffing, 10)

Training, and 11) Drills.

At Callaway Plant the plant response to a fire in specific area that does not require main control room abandonment is that a single operator sequentially performs all recovery actions in a single Fire Response Procedure. The Fire Response Procedure includes NFPA 805 recovery actions in a single fire area. The proceduralized actions in the Fire Response Procedure are the operators only task and he/she would not address anything other than what is explicitly listed in the Fire Response Procedure. In addition, unless stated in the fire procedures, the operator performing the fire response actions would not interact with the main control room. The Operations Technician operates independent from any control room action.

For each fire area all RAs credited for all scenarios in that area were walked down using the Fire Response Procedure and timed with an Operations Technician performing the RAs. The time required to complete each individual RA was conditional on accomplishing all steps prior to the critical steps associated with the RA. The same action in different fire areas could have different completion times based on the complexity of the fire scenarios. Based on the evaluation all RAs were determined to be feasible. Therefore, the feasibility determination included consideration of multiple time critical actions credited for a fire area if they existed. The defense in depth process as a whole ensures there is not over reliance on any one echelon of DID. RAs are included within DID Echelon 3 for consideration.

The Fire Human Reliability Analysis task (FHRA, NUREG/CR-6850 Task 12) evaluated the reliability of each RA given the fire-specific scenario. The FHRA used the results of the feasibly assessment, draft procedure guidance, walkdown information and timeline develop as input to performing the Human Error Probability (HEP) calculation. The FHRA qualitative analysis of the timeline and performance shaping factors accounts for the performance of all actions (DID and NFPA 805 Recovery Actions) included in the Fire Response Procedures. Therefore, the FHRA reliability determination also included consideration of multiple time critical actions credited for a fire area if they existed.

Page 89 of 96 to ULNRC-05851 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.

Response to Probabilistic Risk Assessment RAI 19 This RAI response will be provided with supplemental correspondence.

Page 90 of 96 to ULNRC-05851 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 Use of methods, data, and guidance from EPRI TR-1016735, "Fire PRA Methods Enhancements, Additions, Clarifications, and Refinements to EPRI 1011989," were updated to include any changes as incorporated into Supplement 1 of NUREG/CR-6850 with the following two exceptions:

a. The Callaway Plant Fire PRA did not use FAQ 08-0044 (Chapter 9). Callaway Plant used NUREG/CR-6850 for pump oil fires, which is more conservative. The scenario results did not show undue conservatisms.

b. Chapter 10 (FAQ 08-0048) of NUREG/CR-6850, Supplement 1 provides initiating event frequencies for all initiating event frequency bins. The frequencies from Chapter 10 were used in the Callaway Plant PRA. Chapter 4 (FAQ 06-0017) of NUREG/CR-6850, Supplement 1 provides frequencies and bins for High Energy Arcing faults, but these conflict with the frequencies specified in Chapter 10. The frequencies and bin numbers specified in Chapter 4 of NUREG/CR-6850, Supplement 1 were not used in the Callaway Plant PRA.

Page 91 of 96 to ULNRC-05851 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 The combined PRA Standard Supporting Requirement (SR) PRM-B4 states that any new fire-induced initiating events (IE) that have been identified in accordance with SR PRM-B3 must be modeled in accordance with a number of IE HLRs from section 2 of the ASME standard. The peer review team concluded that the Callaway Plant Fire PRA did not identify any new initiating events in accordance with SR PRM-B3, so that PRM-B4 is not applicable. However, the Callaway Plant PRA did not substantiate why the SR is non-applicable to the Callaway Plant Fire PRA.

The F&O states: to meet this supporting requirement, a defined basis is needed to support the claim of nonapplicability of the requirements.

The resolution for the F&O is to update the Fire-Induced Risk Model Report (Callaway FPRA 17671-004) to explicitly note that no new initiating events have been identified. The draft of the text to be added to 17671-004 is shown below:

SR-PRM-B-4 requires the initiating event analysis for the fire to be consistent with HLR-IE-A, IE-B, and IE-C from part 2 of the PRA standard.

HLR-IE-A requires a thorough and complete identification effort for IEs.

The first step in developing the fire event trees was to review the internal events initiating event candidates for disposition or inclusion in the Fire PRA. This included all IEs which were chosen for the internal events PRA in addition to those that were not chosen. In addition, all IEs from the MSO evaluation in Task 2 were included.

HLR-IE-B requires that if IEs are grouped, it must be determined that all IEs in the group have the same plant response characteristics This HLR is not applicable, because the Fire PRA does not group IEs. Each IE considered in the Fire PRA is explicitly modeled.

HLR-IE-C requires that IE group frequency be accurately determined:

This HLR is not applicable because the IEs are caused by a fire, whose frequency is explicitly calculated in Task 5 of the Fire PRA.

Page 92 of 96 to ULNRC-05851 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.

Response to Probabilistic Risk Assessment RAI 22 This RAI response will be provided with supplemental correspondence.

Page 93 of 96 to ULNRC-05851 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.

Response to Probabilistic Risk Assessment RAI 23 This RAI response will be provided with supplemental correspondence.

Page 94 of 96 to ULNRC-05851 Section 6: Licensee Identified Changes (LIC)

LIC-01: Drawing F/P 095067, Fire Protection System Fire Water Storage Tank General Plan, was added to the reference documents in the LAR Transition Report Attachment A, Table B-1 Section 3.5.2. This drawing provides tank interconnection details which are not available in the previously listed reference drawings.

LIC-02: The following calculations were added to the references for LAR Transition Report Attachment A, Table B-1 Section 3.5.3. These documents provide the specific basis for compliance with this section and supplements the general references previously provided.

x M-KC-316, "Fire Protection System Hydraulic Calculations Determine the Adequacy of the Fire Protection System for Providing the Design Flow and Pressure to the Interface with the Sprinkler System" x M-KC-413, "Fire Protection Determines the Flow Requirements of the Fire Pump,"

were added to the reference document column.

LIC-03: Transition Report Attachment A, Table B-1 Section 3.5.3 x Compliance statement was changed from "Complies with Clarification" to "Complies".

x Compliance Basis statement was changed from "Fire pumps comply with NFPA 20-1974 Edition except as identified below" to "No Additional Clarification".

x The following references were removed from the Reference Document Column:

o Letter ULNRC-00189 from Bryan (UE) to Rusche (NRC) dated April 15, 1977/ Section 9.5.1.1.

o NUREG-0830, "Safety Evaluation Report Related to the Operation of Callaway Plant, Unit No. 1," dated October 1981 / Sections 9.5.1.1 and 9.5.1.6.

o NUREG-1058, "Technical Specifications Callaway Plant, Unit No. 1" /

Sections 4.7.10.1.1.a, 4.7.10.1.1.f, 4.7.10.1.2.a, and 4.7.10.1.2.c.

o A duplicate reference to Calculation KC-27, "NFPA Code Conformance Review," Rev. 0 / Appendix A, Section 20.

o NFPA 20, "Standard for the Installation of Centrifugal Fire Pumps," 1974 Edition / Sections 8-6.1, 8-6.2, and 12-3.1.

LIC-04: The following text regarding system demand was removed from the Compliance Basis column in LAR Transition Report Attachment A, Section 3.5.4. This statement was not needed as our compliance basis is complies via previous NRC approval.

"The greatest water demand for the fixed fire suppression systems is 1300 gpm and, coupled with 500 gpm for hose streams, creates a total water demand of 1800 gpm at Page 95 of 96 to ULNRC-05851 the residual pressure of 80 psig. The staff finds that the water supply system can deliver the required water demand with one pump out of service.

LIC-05: Remove prior approval compliance statement from LAR Transition Report Attachment A, Table B-1 Sections 3.5.3, 3.6.1, and 3.8.1 by removing the prior approval text and associated references regarding non-compliant NFPA code testing frequencies. Test frequencies are addressed by LAR Transition Report Attachment A, Table B-1 section 3.2.3 (1).

LIC-06: Clarified the Compliance Basis description in LAR Transition Report Attachment A, Table B-1 Section 3.6.1. The text was revised to add "in power block buildings" for clarification of the intent of this statement. Deleted text in the compliance basis for Transition Report Attachment A, section 3.6.1 regarding the hose provided at each hose rack as it is not germane to the basis for compliance.

LIC-07: LAR Transition Report Attachment S, Table S-3, item 11-805-061 was revised to correct the listed procedural reference. The reference to APA-ZZ-00701, Control of Fire Protection Impairments, was corrected to APA-ZZ-00703, "Fire Protection Operability Criteria and Surveillance Requirements.

LIC-08: A typographical error was discovered in the LAR Transition Report Table 4-3 Fire Protection Systems and Features. Fire Area A-18 has ERFBS that was credited in the model. The ERFBS is correctly and specifically called out in the FSA; however, Table 4-3 indicates that the ERFBS is not required. Table 4-3 is being updated to indicate ERFBS is required in fire area A-18. This also requires a minor change to the Transition Report Attachment C, Table B-3.

Page 96 of 96 to ULNRC-05851 Page 1 of 7 : Revisions to the Transition Report Main Body to ULNRC-05851 Page 2 of 7 Ameren Missouri Callaway Plant NFPA 805 Transition Report (2) Selection of cables necessary to achieve the nuclear safety performance criteria in Chapter 1 (3) Identification of the location of nuclear safety equipment and cables (4) Assessment of the ability to achieve the nuclear safety performance criteria given a fire in each fire area The NSCA methodology review evaluated the Callaway Plant Calculation KC-26, Nuclear Safety Capability Assessment, Revision 0 methodology against the guidance provided in NEI 00-01, Revision 1 Chapter 3, Deterministic Methodology, as discussed in Appendix B-2 of NEI 04-02. The methodology is depicted in Figure 4-2 and consisted of the following activities:

x Each specific section of NFPA 805 2.4.2 was correlated to the corresponding section of Chapter 3 of NEI 00-01, Revision 1. Based upon the content of the NEI 00-01, Revision 1 methodology statements, a determination was made of the applicability of the section to the station.

x The plant-specific methodology was compared to applicable sections of NEI 00-01, Revision 1 and one of the following alignment statements and its associated basis were assigned to the section:

Aligns Aligns with Intent Not in Alignment Not in Alignment, but Prior NRC Approval Not in Alignment, but no adverse consequences x For those sections that do not align, an assessment was made to determine if the failure to maintain strict alignment with the guidance in NEI 00-01, Revision 1 could have adverse consequences. Since NEI 00-01, Revision 1 is a guidance document, portions of its text could be interpreted as good practice or intended as an example of an efficient means of performing the analyses. If the section has no adverse consequences, these sections of NEI 00-01, Revision 1 can be dispositioned without further review.

(Note: Comparison of the Callaway Plant NSCA methodology (Callaway Plant Calculation KC-26, Nuclear Safety Capability Assessment, Revision 0) to Chapter 3 of NEI 00-01, Revision 1 determined that the methodology aligns with the guidance in applicable sections of NEI 00-01, Revision 1.)

Description of Gap Analysis between NEI 00-01 Revisions 1 and 2:

In May 2009, Revision 2 to NEI 00-01 was issued. Revision 2 in large measure implements the Required for Hot Shutdown (RHSD; Green box) and Important to Safe Shutdown (ISSD; SSA Orange box) criteria for classifying SSD-credited devices and determining the allowable tools RAI 01 to address their failure. In general, Revision 2 incorporates the guidance associated with multiple spurious operations (MSOs) as related to those plants not transitioning to NFPA 805.

Specifically, the methodology in Revision 2 reflects insights gained from, not only the EPRI/NEI Cable Fire Testing, but also the CAROLFIRE Cable Fire Testing, the outcome of meetings with the NRC Staff and information provided within SECY 08-0093 and a draft revision to Regulatory Guide 1.189. These changes were made to address NRC comments related to segregating those components necessary for post-fire hot shutdown (green box, defined in 10CFR50, Appendix R, Section III.G.1.a as one train of systems necessary to achieve and maintain hot shutdown conditions) and those whose mal-operation could provide a potential impact to post-August 2011 Page 15 to ULNRC-05851 Page 3 of 7 Ameren Missouri Callaway Plant NFPA 805 Transition Report ILUHVDIHVKXWGRZQ ³RUDQJHER['GHILQHG&)5$SSHQGL[56HFWLRQ,,,*DVFRPSRQHQWV

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August 2011 Page 16 to ULNRC-05851 Page 4 of 7 Ameren Missouri Callaway Plant NFPA 805 Transition Report Analysis Report (FSAR) Appendix 5.4A Safe Shutdown. Therefore, the unit may remain at or below the temperature defined by a Hot Standby plant operating state.

Results Coping Time The NFPA 805 Nuclear Safety Performance Criteria (NSPC) Analysis for Callaway Plant has been developed to ensure that the plant can achieve and maintain the reactor fuel in a Safe and Stable condition assuming that a fire event occurs during Callaway Plant Mode 1 (Power Operation), Mode 2 (Startup), Mode 3 (Hot Standby), and Mode 4 (Hot Shutdown), up to the point at which the MCC breakers for the Residual Heat Removal Loop Suction Isolation Valves, BBPV8702A, BBPV8702B, EJHV8701A, and EJHV8701B, are unlocked and closed. Refer to Attachment C (Table B-3) for the Systems and Components credited with supporting Safe and Stable plant conditions by fire area.

The NFPA 805 Nuclear Safety Capability Assessment (NSCA) has demonstrated that Callaway Plant can achieve and maintain Safe and Stable conditions for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the minimum shift operating staff before having to take action to recharge the nitrogen accumulators. This initial 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> provides sufficient time for the Emergency Response Organization (ERO) to respond and be available to support Safe and Stable actions to extend Hot Standby conditions.

Coping Time Bases The minimum 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> coping duration is based on the normal operating pressure band of the nitrogen accumulators that support emergency operation of the Steam Generator Atmospheric Steam Dump (ASD) valves and the Turbine Driven Auxiliary Feedwater (TDAFW) Pump to Steam Generator flow control valves. Actions required to sustain Mode 3 (Hot Standby) beyond 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> includes an action to recharge the backup nitrogen accumulator tanks for the ASD valves and the TDAFW Pump to Steam Generator flow control valves. Recharging the tanks requires an operator to open a manual valve in Auxiliary Building fire area A-29. Opening this manual valve is addressed in plant procedures and has been demonstrated to be feasible.

SSA Additionally, the backup nitrogen accumulator tanks may need recharging every 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />

RAI 06

thereafter based on valve cycling demands. Components and/or cables associated with this action are included within the NSCA equipment list.

The ASD valves and TDAFW Pump to Steam Generator flow control valves are air operated with a backup nitrogen gas supply tank. On loss of Instrument Air, which is conservatively assumed for NSCA, the backup nitrogen supply is relied on to maintain valve function from the MCR. The tank capacity is based on an assumed number of valve cycles and initial normal operating pressure. The 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> recharge time is based on the number of valve cycles assumed for a Station Black Out plus the available margin from the lower range of the normal accumulator operating pressure band. Operator action to refill the accumulators can extend the period in which these components can be used.

Impact to Plant if Recharge Time is Exceeded Should the nitrogen accumulator tanks lose adequate pressure inventory the valve function from the MCR would be lost. No damage to the valves would occur and they would retain their capability for full MCR function once the nitrogen tanks are recharged or instrument air recovered. Loss of the ASD function would eventually result in cycling of the steam generator code safety valves. Loss of nitrogen pressure would result in the TDAFW Pump to Steam Generator flow control valves failing open. Flow through these valves can be throttled by a SSA manual valve. Operation in this manner is procedurally controlled and is feasible. RAI 06 August 2011 Page 18 to ULNRC-05851 Page 5 of 7 Ameren Missouri Callaway Plant NFPA 805 Transition Report Methods to Maintain Safe and Stable and Extend Hot Standby Conditions The following describes methods to maintain the Safe and Stable condition and related support actions:

1. Callaway Plant has design features and procedures to ensure that an adequate source of inventory is provided for decay heat removal in sustained Mode 3 (Hot Standby) conditions. If the Condensate Storage Tank inventory is depleted the TDAFW pump suction will automatically transfer to the ESW supply from the Ultimate Heat Sink.

Transfer can be automatic or manual from the Main Control Room.

2. RCS Pressure control is maintained by a combination of ASDs, Pressurizer Heaters, and/or Reactor Pressure Vessel Head Vent valves or PORVs.

3. Core decay heat in Mode 3 (Hot Standby) will be rejected to the secondary plant through one or more of the Steam Generators, and then to atmosphere through the Atmospheric Steam Dump valves.

4. The Callaway Plant reactor core design ensures that Keff is maintained <0.99 while the plant is in sustained Mode 3 (Hot Standby). Gravity insertion of the control rods into the reactor core will ensure reactivity control is achieved for Mode 3 (Hot Standby) for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Subsequently, maintaining Keff <0.99 for Safe and Stable conditions will require boration of the RCS as described in FSAR Appendix 5.4A.

5. Inventory makeup to the RCS may only be required to account for expected RCS leakage and minimal RCS shrinkage as well as RCP seal injection. Callaway Plant has design features and procedures to ensure that an adequate source of borated inventory is provided for RCS inventory control in sustained Mode 3 (Hot Standby) (i.e., RCS inventory makeup from the RWST) utilizing the CVCS system. Callaway Plant has design features and procedures to ensure that an adequate method is provided for RCS inventory control in sustained Mode 3 (Hot Standby) utilizing the Reactor Pressure Vessel Head Vent valves. If RWST inventory is depleted it will be refilled using a combination of Reactor Make Up Water Storage Tank and Boric Acid Storage Tank inventories.

6. Callaway Plant has design features and procedures to ensure that an adequate source of heat input is maintained for RCS pressure control in sustained Mode 3 (Hot Standby)

(i.e., a minimum of 150kW of pressurizer heater input to maintain the RCS sub-cooled) utilizing available combinations of the backup pressurizer heaters (Group A and Group B are 150kW each). The backup pressurizer heaters are capable of being energized from emergency diesel generator power.

7. Each emergency diesel generator (EDG) is provided with a storage tank having a fuel oil capacity sufficient to operate that diesel for a period of 7 days while the EDG is supplying maximum post LOCA load demand discussed in the FSAR, Section 9.5.4.2.

The maximum load demand is calculated based on the fuel consumption by one EDG for operation at continuous rating for 7 days. This onsite fuel oil capacity is sufficient to operate the DGs for longer than the time to replenish the onsite supply from outside sources.

SSA Qualitative Assessment of Risk

RAI 06

The fire brigade will respond to fire events within the Protected Area boundary in accordance with the guidance of EIP-ZZ-00226, Fire Response Procedure For Callaway Plant. If the fire (non-hostile) meets the criteria of EIP-ZZ-00101, Classification of Emergencies, an SSA emergency declaration would be initiated. In the event of an Alert declaration or higher the Shift

RAI 06

August 2011 Page 19 to ULNRC-05851 Page 6 of 7 Ameren Missouri Callaway Plant NFPA 805 Transition Report Emergency Response Organization (ERO) will be supplemented by the On-Site ERO within 30 minutes during normal working hours and within 90 minutes during off-normal hours. The On-Site ERO will assist the Control Room personnel with implementation of the longer term actions necessary to maintain the fuel in a safe and stable configuration.

Following stabilization at Hot Standby, assessment and repair activities would commence to restore plant equipment needed to support RCS cool down in a safe and controlled manner.

ERO resources will be available to assist the MCR in fire damage assessment and restoration of multiple success paths. Note that the Alternate Emergency Power supply (AEPS) is available but not credited in the NSCA.

x The actions required to maintain Safe and Stable conditions are limited.

x Procedures are in place for the Safe and Stable actions identified above.

x The 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> coping period provides reasonable assurance that adequate time is provided for the ERO to be available to augment the minimum plant staffing to support the longer term Safe and Stable actions.

For the most limiting fire scenarios, it is anticipated that the end state of the cool down would be an RCS temperature of approximately 350 F with a long term strategy for reactivity, decay heat removal, and inventory control. Long term subcooled natural circulation decay heat removal is provided by supplying ESW to the Steam Generators and steaming to atmosphere. The extended coping period at these conditions is based on the significant volume of water available for decay heat removal and reduced need for primary make up to match the RCS system losses.

The ERO provides sufficient resources for assessment of fire damage and completion of repairs to equipment necessary to maintain hot standby for an extended period, transition to cold shutdown, or return to power operations as dictated by the plant fire event.

Conclusions The initial coping time is sufficient to allow the ERO to activate. Limited actions are required and procedures are in place for those actions to maintain extended hot standby conditions. The ERO provides adequate capability to extend initial Hot Standby conditions, to transition to cold shutdown, or return to power operations as dictated by the plant fire event. The approach described above has demonstrated the capability to achieve and maintain the reactor fuel in a Safe and Stable condition for an indefinite period following a fire. A qualitative risk assessment has been performed for this scenario which demonstrated that the risk of not being able to SSA maintain the defined safe and stable conditions is acceptably low beyond the defined coping RAI 06 time limit.

Safe and Stable Conditions / Non-Power Operations Assessment interface The Callaway Plant NFPA 805 Non-Power Operations Assessment provides reasonable assurance the reactor fuel is maintained in a safe and stable condition for fires which may occur in Mode 4 (Hot Shutdown) from the point at which the Motor Control Center (MCC) breakers for the Residual Heat Removal Loop Suction Isolation Valves, BBPV8702A, BBPV8702B, EJHV8701A, and EJHV8701B, are unlocked and closed, Mode 5 (Cold Shutdown) and Mode 6 (Refueling). Refer to Section 4.3 for a description of the Callaway Plant Non-Power Operations Assessment for fires that occur in the non-power operational modes.

August 2011 Page 20

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Attachment A to ULNRC-05851 Page 1 of 20 Attachment A: Revision to Transition Report Attachment A

- NEI 04-02 Table B Transition of Fundamental Fire Protection Program and Design Elements

Attachment A to ULNRC-05851 Page 2 of 20 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 Gaseous Hydrogen Systems at Sites," 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

Attachment A to ULNRC-05851 Page 3 of 20 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

Attachment A to ULNRC-05851 Page 4 of 20 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

Attachment A to ULNRC-05851 Page 5 of 20 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

Attachment A to ULNRC-05851 Page 6 of 20 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

Attachment A to ULNRC-05851 Page 7 of 20 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

Attachment A to ULNRC-05851 Page 8 of 20 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

Attachment A to ULNRC-05851 Page 9 of 20 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

Attachment A to ULNRC-05851 Page 10 of 20 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

Attachment A to ULNRC-05851 Page 11 of 20 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

Attachment A to ULNRC-05851 Page 12 of 20 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

Attachment A to ULNRC-05851 Page 13 of 20 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

Attachment A to ULNRC-05851 Page 14 of 20 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

Attachment A to ULNRC-05851 Page 15 of 20 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

Attachment A to ULNRC-05851 Page 16 of 20 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

Attachment A to ULNRC-05851 Page 17 of 20 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

Attachment A to ULNRC-05851 Page 18 of 20 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

Attachment A to ULNRC-05851 Page 19 of 20 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

Attachment A to ULNRC-05851 Page 20 of 20 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

Attachment B to ULNRC-05851 Page 1 of 3 Attachment B: Revisions to Transition Report Attachment B

- NEI 04-02 Table B Nuclear Safety Capability Assessment - Methodology Review

Attachment B to ULNRC-05851 Page 2 of 3 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

Attachment B to ULNRC-05851 Page 3 of 3 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

Attachment C to ULNRC-05851 Page 1 of 7 Attachment C: Revisions to Transition Report Attachment C - NEI 04-02 Table B Fire Area Transition

Attachment C to ULNRC-05851 Page 2 of 7 Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment C - Table B-3 Fire Area Transition Unit Fire Area Description 1 A-18 Electrical Penetration Room A Fire Zone Description 1410 Electrical Penetration Room A Regulatory Basis 4.2.4.2 - Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptions Performance Goal Method of Accomplishment Comments Decay Heat Removal - CSD Use RHR Train B, with cooling water provided by CCW Pumps B and D.

Decay Heat Removal - HSB Steam Generators A and D are supplied by MDAFW Pump B. See VFDR No. A-18-001 and A-18-002 Process Monitoring RCS Pressure Channels I and II Pressurizer Pressure Channel II Pressurizer Level Channel II Ex-core Neutron Monitoring Channel IV RCS Loop A (1) T-cold Temperature Channel II Steam Gen. A Pressure Channel I Steam Gen. A Narrow Range Level Channel IV Steam Gen. A Atmos. Steam Dump Pressure Channel I Aux. Feedwater Flow to Steam Gen. A Channels I and IV RCS Loop D (4) T-hot Temperature Channels II and VI Steam Gen. D Pressure Channel I Steam Gen. D Wide Range Level Channel IV Steam Gen. D Atmos. Steam Dump Pressure Channel IV Aux. Feedwater Flow to Steam Gen. D Channel IV Aux. Feedwater Pump B Suction Pressure Channel IV Aux. Feedwater Pumps Low Suction Pressure (LSP - Auto Transfer to ESW)

Channels I and IV Condensate Storage Tank Level Channel VI Refueling Water Storage Tank Level Channel I Volume Control Tank Level Channels I and IV Core Exit Thermocouples Train B (Channel IV and VI)

RCS Inventory Control Maintain inventory and RCP seal integrity using Charging Pump B via the Boron See VFDR No. A-18-003, A-18-004, A August 2011 Page C-131

Attachment C to ULNRC-05851 Page 3 of 7 Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment C - Table B-3 Fire Area Transition Unit Fire Area Description 1 A-18 Electrical Penetration Room A Injection flowpath and the Alternate RCP Seal Injection flowpath. RCS RV Head 005, and A-18-006 Vent flowpath Train B is available for letdown of RCS inventory, if necessary.

RCS Pressure Control Control pressure using Pressurizer Heater Backup Group B. Use PORV (BBPCV0456A) to depressurize.

Reactivity Control Trip reactor from Control Room. Use Charging Pump B to inject borated water from the RWST.

Vital Auxiliaries Operate CCW Pumps B and D, and ESW Pumps A and B.

Offsite Power to NB01 and NB02 credited.

HVAC credited for Main Control Room and Containment (Train B credited).

Reference Documents Calculation KC-26, Nuclear Safety Capability Assessment Licensing Actions Licensing Action Title Deviation from Section C.5.b of Appendix A to BTP ASB 9.5-1 for Containment Electrical Penetrations Summary Deviation submitted per 3/14/1984 SNUPPS letter to the NRC, justifying non-rated electrical penetrations in the reactor containment walls to Fire Areas A-17 and A-18 NRC in NUREG-0830, Supplement 3, dated 05/1984 based on the following:

1. The containment wall is 4-foot-thick reinforced concrete with a continuous 1/4-inch-thick steel liner.

2. Construction is capable of withstanding a 60-psig overpressure without failure.

3. Penetrations serve special nuclear safety-related purpose.

This deviation is active per Section 9.5.1.2.2.3 of the current FSAR SP. The bases identified, and accepted by the NRC, reflect the current plant configuration and remain valid.

Existing Enginering Equivalency Evaluations (EEEE)

EEEE Title Engineering Evaluation RFR 201009031 Summary Auxiliary steel members that were not protected and represent thermal shorts have been evaluated and were found to have no adverse August 2011 Page C-132

Attachment C to ULNRC-05851 Page 4 of 7 Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment C - Table B-3 Fire Area Transition Unit Fire Area Description 1 A-18 Electrical Penetration Room A effect on the protected structural steel's function in the event of a fire, based on the fire hazards within the fire area.

Variances from Deterministic Requirements (VFDR)

VFDR No. A-18-001 Raceways 1J1097, 1J3A1H, and 4J3C1C are provided with a Darmatt 1-hour rated fire wrap in Fire Zone 1410. Per Callaway CAR 200607577, the 1-hour fire wrap for conduit 1J1097 is degraded (i.e., notched). The fire rating of this ERFBS is degraded from the intended 1-hour rating of the design criteria. Conduit 1J1097 contains one safe shutdown cable, 1ABI20EE, which is the instrument signal cable for ABPT0001. Failure of this cable could cause spurious opening of Steam Generator A Atmospheric Steam Dump Valve ABPV0001. Note that Steam Generator A is credited for Decay Heat Removal in this fire area.

This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a degraded barrier issue.

Disposition The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4, therefore, no further action is required.

VFDR No. A-18-002 ABPV0002-P - Cable damage (2ABI20FE and 2ABI20FH) to Pressure Transmitter ABPT0002. Cable damage can spuriously open the Atmospheric Steam Dump Valve, ABPV0002. The valve is required closed to isolate the main steam pressure boundary for Steam Generator B, to maintain positive control over the rate of RCS cooldown, and to maintain RCS sub-cooling.

Note that Steam Generator B is not credited for Decay Heat Removal in this fire area. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.

Disposition The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4, therefore, no further action is required.

August 2011 Page C-133

Attachment C to ULNRC-05851 Page 5 of 7 Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment C - Table B-3 Fire Area Transition Unit Fire Area Description 1 A-18 Electrical Penetration Room A VFDR No. A-18-003 BBPCV0455A-P - Cable damage (1BBK40AG) to BBPCV0455A. Cable damage can spuriously open the Pressurizer Power Operated Relief Valve, BBPCV0455A (spurious opening is only credible assuming external hot shorts). The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.

Disposition The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4, therefore, no further action is required.

VFDR No. A-18-004 BGHV8149A - Cable damage (5BGK35AB) to BGHV8149A. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice A Outlet Isolation Valve, BGHV8149A. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.

Disposition The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4, therefore, no further action is required.

VFDR No. A-18-005 BGHV8149B - Cable damage (5BGK35BB) to BGHV8149B. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice B Outlet Isolation Valve, BGHV8149B. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.

Disposition The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4, therefore, no further action is required.

August 2011 Page C-134

Attachment C to ULNRC-05851 Page 6 of 7 Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment C - Table B-3 Fire Area Transition Unit Fire Area Description 1 A-18 Electrical Penetration Room A VFDR No. A-18-006 BGHV8149C - Cable damage (5BGK35CB) to BGHV8149C. Cable damage can spuriously open or prevent closure of the Chemical Volume Control System Letdown Orifice C Outlet Isolation Valve, BGHV8149C. The valve is required closed to maintain positive control over RCS Inventory and Pressure. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3. This is a separation issue.

Disposition The VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4, therefore, no further action is required.

Required Fire Protection Systems and Features REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES Required?

Fire Zone Category ID Type Notes S L E R D 1410 Detection 107 Ionization Y N N Y N Detection 114 Ionization Y N N Y N Suppression SKC05 Halon Y N N Y N Suppression SKC35 Wet Pipe Y N N N N electrical chase area only Feature None ERFBS Y N N Y N LIC-08 Legend:

Required?

S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk Evaluation August 2011 Page C-135

Attachment C to ULNRC-05851 Page 7 of 7 Ameren Missouri Callaway Plant NFPA 805 Transition Report Attachment C - Table B-3 Fire Area Transition Unit Fire Area Description 1 A-18 Electrical Penetration Room A Fire Suppression Effects on Nuclear Safety Performance Criteria Halon system actuations are not expected to adversely affect electrical equipment. The effects of moderate energy line break and flooding which include rupture and inadvertent actuations from a Fire Protection System pipe break were evaluated as documented in FSAR Chapter 3. In FSAR Section 3.1.3 it states "Fire fighting systems are designed to assure that their rupture or inadvertent operation will not prevent systems important to safety from performing their design functions. In the areas, drains minimize the potential for flooding damage, such that the standing water would not affect safety-related equipment.

Should manual fire fighting be required, water damage to the electrical equipment in this room could result (with or without associated fire damage); however, the water damage would not adversely affect safe shutdown. Safety related electrical cabinets are mounted on pedestals to protect against water effects and are sealed at the top. Safety related electrical cable in tray is qualified for water exposure. The redundant equipment is located in another fire area. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.

Fire Area Comments None August 2011 Page C-136

Attachment G to ULNRC-05851 Page 1 of 2 Attachment G: Revisions to Transition Report Attachment G - Recovery Actions Transition

Attachment G to ULNRC-05851 Page 2 of 2 Ameren Missouri Callaway Plant NFPA 805 Transition Report In accordance with the guidance provided in NEI 04-02, FAQ 07-0030 Revision 5 (ML110070485), and RG 1.205, the following methodology was used to determine recovery actions required for compliance (i.e., determining the population of post-transition recovery actions). The methodology consisted of the following steps:

x Step 1 - Clear definition of the primary control station(s) and the determination of the pre-transition OMAs that are taken at primary control station(s) (activities that occur in the main control room are not considered pre-transition OMAs). Activities that take place at primary control station(s) or in the main control room are not recovery actions, by definition.

x Step 2 - Determination of the population of recovery actions that are required to resolve VFDRs (to meet the risk acceptance criteria or maintain a sufficient level of defense-in-depth).

x Step 3 - Evaluation of the additional risk presented by the use of recovery actions required to demonstrate the availability of a success path.

x Step 4 - Evaluation of the feasibility of the recovery actions.

x Step 5 - Evaluation of the reliability of the recovery actions.

An overview of these steps and the results of their implementation are provided below.

Step 1 - Clear definition of the primary control station(s) and the determination of pre-transition OMAs that are taken at primary control station(s)

The first task in the process of determining the post-transition population of recovery actions was to apply the NFPA 805 definition of recovery action and the RG 1.205 definition of primary control station to determine those activities that are taken at primary control station(s).

Results of Step 1:

Based on the definition provided in RG 1.205, and the additional guidance provided in FAQ 07-0030 Revision 5 (ML110070485), the following is considered to be the primary control station, with the associated enabling, control, and indication functions as identified:

RP118B, Auxiliary Shutdown Panel PRA

RAI 11

x Enable RP118B with isolation transfer switches/control switches located at RP118B x Steam Generator B (2) pressure indication (ABPIC0002B) x Steam Generator B (2) wide range level indication (AELI0502A) x Steam Generator B (2) AFW flow indication (ALFI0003B) x Open control for steam supply valve from Steam Generator B (2) to TDAFP (ABHV0005) x Open and close control for Steam Generator B (2) Atmospheric Steam Dump Valve (ABPV0002) x Open and close control for Steam Generator B (2) AFW flow control valve from TDAFP (ALHV0010) x Open and close control for Essential Service Water to suction of MDAFW Pump B (ALHV0030)

August 2011 Page G-2

Attachment L to ULNRC-05851 Page 1 of 4 Attachment L: Revisions to Transition Report Attachment L

- NFPA 805 Chapter 3 Requirements for Approval (10 CFR 50.48(c)(2)(vii))

Attachment L to ULNRC-05851 Page 2 of 4 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

Attachment L to ULNRC-05851 Page 3 of 4 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

Attachment L to ULNRC-05851 Page 4 of 4 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

Attachment S to ULNRC-05851 Page 1 of 3 Attachment S: Revisions to the Transition Report Attachment S - Plant Modifications and Items to be Completed During Implementation

Attachment S to ULNRC-05851 Page 2 of 3 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

Attachment S to ULNRC-05851 Page 3 of 3 Ameren Missouri Callaway Plant NFPA 805 Transition Report Table S-3 Implementation Items Item Unit Description LAR Section / Source

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August 2011 Page S-16

Attachment T to ULNRC-05851 Page 1 of 5 Attachment T: Revisions to the Transition Report Attachment T - Clarification of Prior NRC Approvals

Attachment T to ULNRC-05851 Page 2 of 5 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

Attachment T to ULNRC-05851 Page 3 of 5 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

Attachment T to ULNRC-05851 Page 4 of 5 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 analysis and the 100% cited in NUREG-0830, Supplement 3.

05 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

Attachment T to ULNRC-05851 Page 5 of 5 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

Attachment W to ULNRC-05851 Page 1 of 11 Attachment W: Revisions to the Transition Report Attachment W - Fire PRA Insights

Attachment W to ULNRC-05851 Page 2 of 11 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

Attachment W to ULNRC-05851 Page 3 of 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 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.

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Attachment W to ULNRC-05851 Page 4 of 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 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.

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Attachment W to ULNRC-05851 Page 5 of 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 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.

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Attachment W to ULNRC-05851 Page 6 of 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 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.

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Attachment W to ULNRC-05851 Page 7 of 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 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.

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Attachment W to ULNRC-05851 Page 8 of 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 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.

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Attachment W to ULNRC-05851 Page 9 of 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 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.

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Attachment W to ULNRC-05851 Page 10 of 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 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.

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Attachment W to ULNRC-05851 Page 11 of 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 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.

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