ULNRC-05876, Callaway, Unit 1 - Response to Request for Additional Information Re Adoption of National Fire Protection Association Standard NFPA 805: Difference between revisions

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{{Adams
{{Adams
| number = ML12194A638
| number = ML12194A624
| issue date = 07/12/2012
| issue date = 07/12/2012
| title = Callaway, Unit 1 - Response to Request for Additional Information Re Adoption of National Fire Protection Association Standard NFPA 805
| title = Callaway, Unit 1 - Response to Request for Additional Information Re Adoption of National Fire Protection Association Standard NFPA 805
| author name =  
| author name = Maglio S
| author affiliation = Ameren Missouri, Union Electric Co
| author affiliation = Ameren Missouri, Union Electric Co
| addressee name =  
| addressee name =  
| addressee affiliation = NRC/NRR
| addressee affiliation = NRC/NRR, NRC/Document Control Desk
| docket = 05000483
| docket = 05000483
| license number =  
| license number =  
| contact person =  
| contact person =  
| case reference number = ULNRC-05876
| case reference number = ULNRC-05876
| package number = ML121940471
| package number = ML121940741
| document type = - No Document Type Applies
| document type = Letter type:ULNRC
| page count = 237
| page count = 4
}}
}}


=Text=
=Text=
{{#Wiki_filter:Enclosure 1 to ULNRC-05876 Page 1 of 136 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
{{#Wiki_filter:WAmeren MISSOURI July 12, 2012 ULNRC-05876 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Ladies and Gentlemen:
 
10 CFR 50.90 DOCKET NUMBER 50-483 CALLA WAY PLANT UNIT 1 UNION ELECTRIC CO. FACILITY OPERATING LICENSE NPF -30 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING ADOPTION OF NATIONAL FIRE PROTECTION ASSOCIATION STANDARD NFPA 805 Callaway Plant Reference 1 ): Ameren Missouri Letter ULNRC-05851, Response to Request for Additional Information Regarding Adoption ofNational Fire Protection Association Standard NFP A 805 dated April 17, 2012 (T AC No. ME7046) On March 2, 2012, Callaway Plant formally received from the NRC Staff a request for additional information (RAI) related to the application request to adopt a new fire protection licensing basis based on NFPA 805 which complies with the requirements in 10 CFR 50.48(a) and (c) and the guidance in Revision 1 of Regulatory Guide (RG) 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants". On April17, 2012 Ameren Missouri transmitted to the NRC Staff via Reference 1, responses to a portion of the RAI questions.
Section 3:  Response to Monitoring Program RAIs
On April 12th, 17th, & June 21th, 2012, teleconferences between Ameren Missouri and the NRC staff reviewers were conducted in which additional requests/questions on the remaining RAI's were discussed.
 
Following these teleconferences the NRC provided to Ameren Missouri supplemental requests/questions to the original RAI's. The purpose of this letter is to provide the additional responses to the RAI transmittal dated March 2, 2012 including responses to the supplemental requests/questions identified during the teleconferences.
Section 4:  Response to Safe Shutdown RAIs
PO Box 620 Fulton, MD 65251 AmerenMissouri.com ULNRC-05876 July 12, 2012 Page2 Enclosure 1 provides the RAI responses not provided in Reference
 
: 1. Enclosure 1 also contains a description of changes to the license amendment request (LAR) that Ameren Missouri has identified as being warranted.
Section 5:  Response to Probabilistic Risk Assessment RAIs Section 6:  Licensee Identified Ch anges to the Transition Report
Attachments are provided within the enclosure to show the LAR Transition Report changes from the original submittal.
  :  Revisions to the Transition Report Main Body 
Enclosure 2 provides the computer program input files requested in Fire Modeling RAis 2b, 3b and 3t. This submittal does contain revised/new commitments which are provided in the revised LAR Transition Report Attachment S, Plant Modifications and Items to be Completed during Implementation.
 
The No Significant Hazards Consideration determination provided in the original submittal is not altered by the additional information provided in this letter. If there are any questions, please contact Mr. Scott Maglio at (573) 676-8719 or Mr. Roger Wink at (314) 225-1561.
Attachment A:  Revisions to Transition Report Attachment A - NEI 04-02 Ta ble B Transition of Fundamental Fire Protection Program and Design Elements Attachment B:  Revisions to Transition Report Attachment B - NEI 04-02 Table B Nuclear Safety Capability Assessment - Methodology Review Attachment C:  Revisions to Transition Report Attachment C - NEI 04-02 Table B Fire Area Transition Attachment D:  Revisions to Transition Report Attachment D - Non-Power Operational Modes Transition Attachment E:  Not used.
I declare under penalty of peijury that the foregoing is true and correct. Executed on: '1 ftL-I U 11..-.RCW/nls Sincerely, Regulatory Affairs Manager Enclosure 1, Request for Additional Information (RAI) with Callaway Plant Response Enclosure 2, Computer input files for Fire Modeling RAis 2b, 3b and 3t.
Attachment F:  Not used.
ULNRC-05876 July 12, 2012 Page 3 cc: U.S. Nuclear Regulatory Commission (Original and 1 copy) Attn: Document Control Desk Washington, DC 20555-0001 Mr. Elmo E. Collins Regional Administrator U.S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. Andrew Hon Senior Project Manager, Callaway Plant Office ofNuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-8G 14 Washington, DC 20555-2738 ULNRC-05876 July 12, 2012 Page4 Index and send hardcopy to QA File A160.0761 Hardcopy:
Attachment G: Not used Attachment H:  Revisions to Transition Report Attachment H - NFPA 805 Frequently Asked Question Summary Table to ULNRC-05876
Certrec Corporation 4150 International Plaza Suite 820 Fort Worth, TX 76109 (Certrec receives ALL attachments as long as they are non-safeguards and may be publicly disclosed.)
 
Electronic distribution for the following can be made via Tech Spec ULNRC Distribution:
Page 2 of 136 Attachment I:  Not used.
A. C. Heflin F. M. Diya C. 0. Reasoner III D. W. Neterer L. H. Graessle J. S. Geyer S. A. Maglio R. Holmes-Bobo NSRB Secretary T. B. Elwood Mr. Bill Muilenburg (WCNOC) Mr. Tim Hope (Luminant Power) Mr. Ron Barnes (APS) Mr. Tom Baldwin (PG&E) Mr. Mike Murray (STPNOC) Ms. Linda Conklin (SCE) Mr. John O'Neill (Pillsbury Winthrop Shaw Pittman LLP) Missouri Public Service Commission Mr. Dru Buntin (DNR)}}
Attachment J:  Revisions to Transition Report Attachment J - Fire Modeling V&V
 
Attachment K:  Not used.
Attachment L:  Revisions to Transition Report Attachment L - NFPA 805 Chapter 3 Requirements for Approval (10 CFR 50.48(c)(2)(vii))
Attachment M:  Not used.
Attachment N:  Not used.
Attachment O: Not used.
Attachment P:  Not used.
Attachment Q:  Not used.
Attachment R:  Not used.
 
Attachment S:  Revisions to Transition Report Attachment S - Plant Modifications and Items to be completed during Implementation Attachment T:  Revisions to Transition Report Attachment T - Clarification of Prior NRC Approvals 
 
Attachment U:  Not used.
Attachment V:  Revisions to Transition Report Attachment V - Fire PRA Quality Attachment W:  Revisions to Transition Report Attachment W
- Fire PRA Insights
 
Attachment X: Revisions to Transition Report Attachment X - Other Requests for Approval to ULNRC-05876  
 
Page 3 of 136 Section 1: Response to Fire Modeling RAIs Fire Modeling RAI 01 Section 2.7.3.2, "Verification and Validation," of NFPA 805 states: "Each calculational model or numerical method used shall be verified and validated through comparison to test results or comparison to other acceptable models."  Section 4.5.1.2 of the Transition Report of the LAR states that a fire m odeling 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 Met hods for the U.S. Nuclear Regulatory Commission Fire Protection Inspection Program," December 2004 (ADAMS Accession No. ML043290075) (FDTs) or NUREG-1824, "Verificat ion &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 veri fied (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 NU REG-1824," and that "the correlation is used within the limits of its range of applicability."
* Flame Height (Method of Heskestad)
* Plume Centerline Temperatur e (Method of Heskestad)
* Radiant Heat Flux (Point Source Method)
* Hot Gas Layer (Method of MQH)
* Hot Gas Layer (Method of Beyler)
* Hot Gas Layer (Method of Foote, Pagni, and Alvares [FPA])
* Hot Gas Layer (Method of Deal and Beyler)
* Ceiling Jet Temperature (Method of Alpert)
* Smoke Detection Actuation Correlation (M ethod of Heskestad and Delichatsios)
The fact that a correlation is used within its range of applicability does not guarantee that it is applied within the validated range reported in NUREG-1824, "V erification and Validation of Selected Fire Models for Nuclear Power Plant Applications," May 2007 (ADAMS Accession    to ULNRC-05876
 
Page 4 of 136 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 re ported in NUREG-1824.
Additional Clarific ation Needed To demonstrate that the models are applied within the range of applicability, the normalized parameters described in NUREG 1824 and 1934 were calculated and the calculations showed that the parameter is within the validated range or justification wa s provided for using a parameter value outside the validated range.
* Fire Froude Number: Explain in more detail how you determined quantitatively, the effect of using a conservative convection fraction of 0.7 on the safety margin in the plume ZOI calculation. In addi tion, describe the criteria that were used to judge that the margin of safety is sufficient.
* Ceiling Jet Radial Distance Relative to Ceiling Height: Provide the normalized parameter values and discuss in light of the sprinkler activation calculations.
* Compartment Aspect Ratio: Confirm that the correct damage temperature in each of the fire zones where the sensitivity analysis was conducted is in fact 330 °C and not 205
°C.
For the case of fires which postulate propagation to secondary combustibles and a Froude number above the validation range, the licensee i ndicates "the fire conditions are calculated assuming a nominal base area which is smaller th an the area of the ignition source."  Provide examples where this assumption was made.
: c. Attachment J of the Transition Report states that the following models are verified and validated on the basis that they are described in an authoritative publicat ion in fire protection literature:
* Heat Detection Actuation Correlation
* Sprinkler Activation Correlation
* Corner and Wall Heat Release Rate
* Correlation for Heat Release Rate s of Cables (Method of Lee)
* Correlation for Flame Spread over Horizontal Cable Trays (FLASH-CAT)
Furthermore, the Transition Report states that these models are used within their range of applicability, which does not guarantee that they are applied within the validated range. Please provide technical details to demonstrate that the model has been applied within its validated range or to justify the application of the model outside its validated range.
 
Additional Clarific ation Needed To demonstrate that the models are applied within the range of applicability the normalized parameters described in NUREG 1824 and 1934 were calculated a nd either showed that the    to ULNRC-05876
 
Page 5 of 136 parameter is within the validated range or ju stification was provided for using a parameter value outside the validated range.
* Sprinkler Activation Correlation: Explain in more detail how the Ceiling Jet Distance Ratio normalized parameter was calculated.
The draft response describes the r/H calculation used in the sensitivity analysis. Is the H equal to the distance between the floor and ceiling or the distance between the ignition source and the ceiling?  In addition, for Fire Area A-16, explain in more detail how the 'critica l heat release rate for sprinkler activation' was calculated in FDT 10.
 
Indicate exactly where in NUREG 1824 it is stated that the total heat release rate was used in the validation of Alpert
's ceiling jet correlation.
: d. Attachment J of the Transition Report states that the "Plume Radius (Method of Heskestad) model is verified and validated on the basis that it is described in an authoritative publication in the fire protection literature. Please provide technical details to demonstrate that the model has been applied within its validated range or to justify the application of the model outside its validated range.
: e. Attachment J of the Transition Report states that the verification and validation of the following applications of Fire Dynamics Simulator (FDS) are documented in NUREG-1824.
* Hot Gas Layer (HGL) Calculations using FDS
* Sprinkler Actuation Ca lculation using FDS
* Temperature Sensitive Equipment Zone of Influence Study using FDS
* Plume/Hot Gas Layer Interaction Study using FDS Please provide technical documentation that demonstrates that FDS was either used within the range of its validity as described in NUREG
-1824 or that the use of FDS outside the verification and validation range in NUREG-1824 is justified.
Additional Documentation Required Staff needs to review the report of the sensitivity study mentioned on page 11 of the draft response to confirm that it is acceptable for the Radial Distance Relative to Fire Diameter to be outside the validated range. The sensitivity analysis report needs to be posted on the portal.
: f. Attachment J of the Transition Report states that the verification and validation of the following applications of Consolidated Mode l of Fire and Smoke Transport (CFAST) are documented in NUREG-1824.
* HGL Calculations using CFAST (Version 6)
* Temperature Sensitive Equipment Hot Gas Layer Study using CFAST
* Control Room Abandonment Calculation using CFAST to ULNRC-05876
 
Page 6 of 136 Please provide technical documentation that demonstrates that CFAST was either used within the range of its validity as described in NUR EG-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 th e CFAST calculation described on page J-6 of the Transition Report were not listed as one of the fire models utilized in the application in Section 4.5.1.2.
Additional Clarification/
Documentation Required
* MCR Study: The justification for the flame length ratio normalized parameter is adequate, however, there is a confusing comment about the ACRS recently deciding that this parameter should be based not only on the flame height, but the height of the fire above the
 
floor and that this is different from NUREG 1934. This is confusing, since the latest version of NUREG 1934 does include the base height as well as flame height in this normalized parameter. Review this last justification statement a nd consider whether it should be amended or stricken from the response.
* In addition, in the justificati on for the equivalence ratio (natural ventilation) normalized parameter being outside of the validation range for heat release rate s higher than 312 kW, it is mentioned that a sensitivity study may be warranted, but none is provided in the response. Since the natural ventilation cases with the highest heat release rate bins seem to be the worst-case in terms of the calculated evacuation times, it is requested that this sensitivity study be provided.
Provide the results of the sensi tivity study related to e quivalence ratio on th e portal or, if those results are already posted, indicate where they can be found.
: g. During the audit, the NRC staff observed that part of the fire modeling performed in support of the transition to NFPA 805 is described in Engineering Planning & Management, Inc. (EPM)
Report No. R1984-001-002, "Callaway Plant Verifi cation and Validation of Fire Modeling
 
Tools and Approaches." Appendices B, C, and D of this report describe FDS and CFAST fire modeling studies of plume/HGL interaction, temperature sensitive equipment zone of influence (ZOI) and HGL effects. Please provide the basis of assurance that the use of the conclusions from these studies in subsequent fire modeling analysis was within the limits of applicability.
Statement and Justification Needed If the results of the studies described in Appendices B, C and D of the V&V report were used in the analyses within their limits of applicability, a statemen t should be provided indicating so. If not, a statement should be provided indi cating why not. In addition, the plume-HGL interaction study in Appendix B is based on calculations for a single ambient temperature (70°F), heat release rate (211 kW) and fire size (physical dimensions) and height. Provide justification for drawing genera lized conclusions based solely on calculations for these input parameter values. to ULNRC-05876
 
Page 7 of 136 Additional Justification Needed Provide a list of areas and scenarios where the results of the anal yses described in Appendix B, C and D of document EPM R1984-01-002 were used.
: h. Section 4.5.1.2 of the Transition Report lists "Multi-Compartment Analysis Hot Gas Layer Analysis" as one of the fire models utilized in the application. However, there is no verification and validation basis provided for this model in Attachment J. Please explain where this fire model was utilized in the application (if a pplicable) 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 va lidation 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 th at the use of FDS outside the verification and
 
validation range in NUREG-1824 is justified. In addition, please explain why this particular calculation was not listed in Section 4.5.1.2 or Attachment J of the Transition Report.
 
Additional Justification Needed Reference is made in the draft response to NIST GCR 07-911 for V&
V of Cleary's smoke detector algorithm, which is now implemented in FDS. The validation in this NIST report is based on three sets of experiments: UL 217 Tests, Room- Corridor-Room Fire Tests and NIST 'Performance of Home Smoke Alarms' Test Validation. 
: i. UL 217 Tests: Page 24 of the NIST report states that the smoke (soot) yield of polystyrene was estimated as the average of the values in the literature for polys tyrene and styrene.
Ignoring the fact that the use of the average in itself is questiona ble, the SFPE Handbook gives soot yields for polysty rene and styrene as 0.135 a nd 0.177 g/g, respectively, leading an average of 0.156 g/g. This is approximately three times what was used in the analysis for fire area C-21.
ii. Room-Corridor-Room Fire Test Validation: These experiments consisted of pool fire experiments with a mix of 75% heptane a nd 25% toluene. The SFPE Handbook lists soot yields for heptane and toluene (listed directly after benzene, but the draft response indicated it was not available) as 0.037 and 0.178 g/g, respectively. The valid ated soot yield used in these experiments was approximately 0.072 g/g, wh ich is a little bit hi gher than the value used in the C-21 analysis (0.05 g/g).
to ULNRC-05876
 
Page 8 of 136 iii. NIST 'Performance of Home Smoke Alarms' Test Validation: As mentioned in the draft response, there is no validated range in this study that can be compared to the analysis in C-
: 21. In addition, the draft response has some discussion about the liter ature values of soot yield for polystyrene mattresses and gives the range of 0.056-0.227 g/g. However, the low and high limits of this range are for polyethylen e and polyurethane, not polystyrene. The SFPE Handbook gives a range of soot yield for polystyrene foam of 0.18 - 0.21 g/g, which is much higher than was used in the analys is for fire area C-21. Provide additional justification for the validity of Cleary's algorithm implemented in FDS for a smoke yield of 0.05 g/g. In addition, explain how the use of a soot yield of 0.05 g/g is justified.
 
Provide the fuel composition that was used in the FDS analyses of areas C-21 and C-22 and explain how this composition was determined.
: j. During the audit, the NRC staff observed th at the software package PyroSim (Version 2010.1.0928) was used to build the FDS input files. Please provide technical documentation that demonstrates that PyroSim is verified to build the input file correctly.
Additional Documentation Required. 
 
The draft response states that there is a discussion in th e FDS report (R1984-001-001) about the verification of PyroSim. However, PyroSim is not mentioned in the report that is on the portal. Provide the revised FDS report on the portal that discu sses PyroSim or explain this discrepancy.
Response to Fire Modeling RAI-01 a) Response provided by ULNRC-05851 dated April 17, 2012.
b) In most cases, the subject correlations have been applied within the validated range reported in NUREG-1824, "Verification and Valid ation of Selected Fire Mode ls for Nuclear Power Plant Applications," Final Report, April 2007. In cases where the models have been applied outside the validated range reported in NUREG-1824, these have been justified as acceptable, either
 
by qualitative analysis, or by quantitative sensitiv ity analysis. Technical details demonstrating the models are within range, as well as any justification of models outside the range, have been documented in Report No. R1984-001-002, "Verification and Validation of Fire Modeling Tools and Approaches for Use in NFPA 805 and Fire PRA Applications,".
c) Refer to the individual bulleted response items below:
* Heat Detection Actuation Correlation The NUREG-1805, "Fire Dynamics Tools: Quantitative Fire Hazard Analysis Methods for the US Nuclear Regulatory Commission Fire Protection Inspection Program," December 2004, heat detection actuation correlation was not considered for fire modeling at    to ULNRC-05876
 
Page 9 of 136 Callaway Plant. There are limited quantities of heat detectors installed at Callaway Plant. Only those heat detectors near the Turbine Generator were credited in the detailed fire modeling. These heat detectors we re credited as part of the pre-action suppression system, which is required for the use of the conditional probability of a catastrophic Turbine Generator scenario found in the guidance of NUREG/CR-6850, "Fire PRA Methodology
 
for Nuclear Power Facilities," Final Report, September 2005, Appendix O. In following this guidance, suppression and detection timing are not required and, therefore, the heat
 
detection actuation correlation wa s not considered. This correlation has been deleted from Section 4.5.1.2 and Attachment J of the Transition Report was revised as shown in Attachment 1 and Attachment J to this enclosure.
* Sprinkler Activation Correlation The Sprinkler Activation Correlation uses the Alpert ceiling jet correlation in addition to a correlation that accounts for the time required to heat the thermal link of the sprinkler. The Alpert ceiling jet correlation is validated in NUREG-1824, "Ver ification and Validation of Selected Fire Models for Nuclear Power Plant Applications," Final Report, April 2007. In
 
most cases, the sprinkler activation correlation has been applied within the validated range reported in NUREG-1824. In cases where the correlation has been applied outside the validated range reported in NUREG-1824, it has been justified as acceptable, either by qualitative analysis, or by quantitative sensitivity analysis. Technical details demonstrating the correlation is used within the validated range, as well as any justification of models outside the range, have been documented in Report No. R1984-001-002, "Verification and Validation of Fire Modeling Tools and Appr oaches for Use in NFPA 805 and Fire PRA Applications,".
* Corner and Wall Heat Release Rate The Corner and Wall Heat Release Rate correlation is applied within the validated range reported in the studies of Zukoski 1, Sargent 2 , Cetegen 3 and Williamson 4 or has been justified as acceptable by qualitative analysis or quantitative sensitivity analysis. Technical details have been documented in Report No. R1984-001-002, "Verification and Validation of Fire Modeling Tools and Approaches for Use in NFPA 805 and Fire PRA Applications,".
 
1 Zukoski, E.E., "Properties of Fire Plumes," Combustion Fundamentals of Fire, Cox, G., Ed., Academic Press, London, 1995.
2 Sargent, W.S., "Natural Convection Flows and Associated Heat Transfer Processes in Room Fires," Ph.D. thesis, California Institute of Technology, Pasadena, CA, 1983.
3 Cetegen, B.M., "Entrainment and Flame Geometry of Fire Plumes," Ph.D. thesis, California Institute of Technology, Pasadena, CA, 1982.
4 Williamson, R.B. Revenaugh, A. and Mowrer, F.W., "Ignition Sources in Room Fire Tests and Some Implications for Flame Spread Evaluation," International Association of Fire Safety Science, Proceedings of the Third International Symposium, New York, pp. 657-666, 1991.
to ULNRC-05876
 
Page 10 of 136
* Correlation for Heat Release Rate s of Cables (Method of Lee)
The Correlation for Heat Release Rates of Cables (Method of Lee) is applied to configurations similar to those reported in NBISR 85-3195, "Heat Release Rate Characteristics of Some Combus tible Fuel Sources in Nuclear Power Plants," July 1985 or has been justified as acceptable by qualitative analysis. Technical details have been documented in Report No. R1984-001-002, "Verifi cation and Validation of Fire Modeling Tools and Approaches for Use in N FPA 805 and Fire PRA Applications,".
* Correlation for Flame Spread over Horizontal Cable Trays (FLASH-CAT)
The Correlation for Flame Spread over Horizontal Cable Trays (FLASH-CAT) model is applied to configurations similar to t hose reported in NUREG/CR-7010, "Cable Heat Release, Ignition, and Spread in Tray Inst allations During Fire (CHRISTIFIRE)," Draft Report for Comment, September 2010 or has been justified as acceptable by qualitative analysis. Technical details have been documented in Report No. R1984-001-002, " Verification and Validation of Fire Modeling Tools and Approaches for Use in NFPA 805 and Fire PRA Applications,".
d) The Heskestad plume radius correlation used for fire modeling at Callaway Plant is the same that is used in EPRI fire model FIVE-Rev1, and that is documented in the SFPE Handbook of Fire Protection Engineering, 4th edition, Secti on 2, Chapter 1. This correlation was not specifically verified and validated (V&V'd) in NUREG-1824, however, Page 2-7 of the 4th edition of the SFPE Handbook of Fire Protection E ngineering states that the value calculated by this correlation is the point where the temper ature has declined to half of the centerline plume temperature. The Heskestad centerline plume correlation was V&V'd in NUREG-1824.
 
The plume radius correlation was used in Engineering, Planning and Management Inc. (EPM) Fire Modeling Database (FMDB) to approximate when to apply the vertical fire plume zone of influence (ZOI), versus the horizontal heat flux based ZOI. The plume radius was not used as the sole basis for any target failures, nor was it used to estimate target temperature. In other words, targets located within the plume radius we re considered to be exposed to the centerline temperatures of the plume, while targets located beyond the plume radius were considered to be exposed to the heat flux as determined by the point source model. 
 
Based on how the  plume radius was applied, and since the plume radius correlation is a derivative of the Heskestad centerline plume temperature correlation, which was V&V'd by NUREG-1824, the plume radius correlation is subject to the same validated ranges which are described in Callaway Fi re Modeling RAI 01-b.
e) In most cases, the FDS analyses have been utilized within the validated range reported in NUREG-1824, "Verification and Valid ation of Selected Fire Mode ls for Nuclear Power Plant Applications," Final Report, April 2007. In cases where the models have been applied outside the validated range reported in NUREG-1824, these have been justified as acceptable, either
 
by qualitative analysis, or by quantitative sensitiv ity analysis. Technical details demonstrating    to ULNRC-05876
 
Page 11 of 136 the models are within range, as well as any justification of models outside the range, have been documented in R1984-001-002, "Verification and Validation of Fire Modeling Tools and Approaches for Use in NFPA 805 and Fire PRA Applications," Revision 1.
f) HGL Calculations using CFAST (Version 6)
 
Hot gas layer calculations at Callaway Plant were not performed using CFAST. Hot gas layer calculations were performed using the NUREG-1805, "Fire Dynamics Tools: Quantitative Fire Hazard Analysis Methods for the US Nuclear Regulatory Comm ission Fire Protection Inspection Program," December 2004 and the National Institute of Standards and Technology (NIST) Fire Dynamics Simulator (FDS). 
 
The HGL Calculations using CFAST model has been deleted from Attachment J of the Transition Report as shown in Attachment J to this enclosure.
 
Control Room Abandonment Calculation using CFAST and Temperature Sensitive Equipment Hot Gas Layer Study using CFAST 
 
In most cases, the subject models have been applied within the validat ed range reported in NUREG-1824, "Verification and Valid ation of Selected Fire Mode ls for Nuclear Power Plant Applications," Final Report, April 2007. In cases where the models have been applied outside the validated range reported in NUREG-1824, thes e have been justified as acceptable, either by qualitative analysis, or by quantitative sensitivity analysis. 
 
Technical details demonstrating the Temperature Sensitive Equipment Hot Gas Layer Study is within the validated range, as well as any justification of models outside the range, have been documented in Report No. R1984-001-002, " Verifi cation and Validation of Fire Modeling Tools and Approaches for Use in N FPA 805 and Fire PRA Applications,".
 
Technical details demonstrating the Control Room abandonment models are within range, as well as any justification of mode ls outside the range, will be documented in the next update to Callaway Plant calculation 17671-010b, "Callawa y NFPA 805 Fire PRA - Main Control Room Fire Analysis" per Implementation Item 12-805-002 as shown in the updated Attachment S to this enclosure.
g) The conclusions from the studies in Appendi ces B, C, and D of R1984-001-002, "Callaway Plant Verification and Validation of Fire Modeling Tools and A pproaches" were used within the limits of applicabil ity established within the study. A detailed discussion of each appendix is provided below:
 
The conclusions from the plume/HGL interaction study in Appendix B of R1984-001-002 were applied in the fire modeli ng analysis by correlating each modeled fire compartment to a generic test category analyzed in Appendix B. The correlation was made between fire compartment parameters (i.e., compartment volum e and ceiling height), w ith consideration of fire scenario characteristics (i.e., heat release rate, fire size and elevation). Fire compartments    to ULNRC-05876
 
Page 12 of 136 with parameters within the limits of a generic test category were judged to perform similarly with respect to potential plum e and hot gas layer interact ion. Section 7.7.1, "Basis for Selection of Fire Modeling Tools," within each fire compartment-specific detailed fire modeling report, discusses the details of how the results of the plum e/HGL interaction study were applied.
 
The objective of the Appendix B study was to establish a baseline for whether limitations of using the Heskestad centerline plume temperatur e correlation calculati ons, with respect to HGL interaction, need to be accounted for. As such, certain parameters were held constant and this is justified as follows:
* Ambient Temperature. The study aims to find a deviation between the estimated plume temperature with and without a hot gas layer. This deviation is strictly a delta temperature. Since the delta temperature is the item of interest, the ambient temperature selected is not relevant to the analysis.
* Heat Release Rate. The heat release rate selected represents the ma jority of fixed ignition sources at Callaway Plant. For those fire compartments where a larger initiating fire is possible or where secondary combustibles may be impacted, the fire scenarios were evaluated, more specifically, to ensure that the target damage set selected bounds any plume/HGL effects.
* Fire Size and Elevation. The fire elevation selected represents a typical electrical cabinet, and bounds the majority of fixed ignition sources at Callaway Plant. For those ignition sources with a fire size (physical dimensions) or elevation exceeding that which was analyzed in Appendix B, the fire scenarios were evaluated, more specifically, to ensure that the target damage set selected bounds any plume/HGL effects.
to ULNRC-05876
 
Page 13 of 136 Examples of the application of Appendix B are as follows: Fire Area Fire Zone Zone Height (ft) Zone Volume (ft 3) Appendix B Category Treatment of Fire Sources in the Model Fixed Ignition Source Fires Transient Fires A-8 1316 13 2,080 Category III There are no fixed ignition sources in this fire zone All targets to the ceiling are considered damaged. Therefore the plume-HGL interaction is bounded by the target damage set. TB-1 3705 12 4,800 Category III Fire scenarios damage all targets to the ceiling; therefore the plume-HGL interaction is bounded by the target damage set. A-22 1512 14 41,580 Category II For all ignition sources that are able to damage targets (i.e., not well sealed and robustly secured), target damage is taken to the ceiling. Therefore, the plume-HGL interaction is bounded by the target set. All targets to the ceiling are considered damaged. Therefore the plume-HGL interaction is bounded by the target damage set. TB-1 3619 14 14,336 Category II This fire zone is open to Fire Compartment TB-1. Due to the large volume in this fire area, no hot gas layer will form.
C-9 3301 15.2 47,630 Category I Fires in this room are bounded by (1) the large open geometry that limits plume/HGL interaction, (2) targets damaged to the ceiling, (3) a HRR less than 55kW and/or (4) the scenario represents suppression activation prior to 20 minutes credited to prevent damage to fully enclosed cable trays only.
The transient fire size postulated in this room is a 69kW fire. Transient scenarios are not large enough to impact secondary combustibles. Due to the large size of the room and the small initiating fire, there will be no hot gas layer and no plume/HGL interaction caused by the transient scenarios. TB-1 4351 17.7 173,184 Category I This large fire zone is also open to Fire Compartment TB-1. Due to the large volume in this fire area, no hot gas layer will form.
The conclusions from the temperature sensitive equipment zone of influence (ZOI) study in Appendix C of R1984-001-002 were applied to the radiant heat ZOI of temperat ure sensitive components housed in vented metal cabinets. Field conditions were verified to ensure the temperature sensitive equipment met the assumptions and limits of the study (i.e., similar cabinet dimensions and construction, minimum venting requirements at the top and bottom of the cabinet). Section 8.3, "Impacts on Sensitive Equipment," within each fire compartment-specific detailed fire modeling report, discusses the details of how the results of the temperature sensitive equipment zone of influence study were applied. Examples of the application of Appendix C are as follows:
Transient Scenarios D-1.TS-06, D-1.TS-07 and D-1.TS-08 were postulated to surround panels KJ121, NE107 and NG03D, respectively. The panels were confirmed to be vented via field walkdowns. The transient foot print was develo ped by adding 3-ft along the length and width of the panel and subtracting the percentage of the area occupied by the equipment. This transient footprints are illustrated in Attachment 3 to Calculation KC-75 "Detailed Fire Modeling Report for Fire Compartment D-1," Revision 0. The calculation of the transient footprint area is shown in Attachment 7 to KC-75. The conclusions from the temperature sensitive equipment hot gas layer study in Appendix D of R1984-001-002 were applied in the fire modeling analysis by correlating each modeled fire compartment to a generic test category analyzed in Appendix D. The correlation was made between    to ULNRC-05876
 
Page 14 of 136 fire compartment parameters (i.e., compartment volum e and ceiling height), with consideration of fire scenario characteristics (i.e., heat release rate and fire growth profile). Fire compartments with parameters within the limits of a generic test category were judged to perform similarly with respect to gas layer formation. Section 8.3, "Impacts on Sensitive Equipment," within each fire compartment-specific detailed fire modeling report, discusses the details of how the results of the temperature sensitive equipment hot gas layer study were applied. Examples of the application of Appendix D are as follows: Fire Area Fire Zone Zone Height (ft) Zone Area (ft 2) Appendix D Category Application of Appendix D C-16 3409 15 354 IV Based on the configuration of the room, there is a possibility for the postulated fires to generate a hot gas layer capable of damaging all equipment and cable targets in the room. The analysis therefore assumes whole room damage for all fire scenarios in this fire zone.
A-1 1102 24 1000 III This fire zone contains no Fire PRA equipment targets and is open to Fire Area A-1. Therefore, there is no impact of hot gas layer on temperature sensitive equipment for the fire scenarios in this fire zone.
C-9 3301 15 3139 II Based on the results of the study, a hot gas layer is not expected to descend to the height of the plant equipment. Only the tops of the equipment in the area are likely to be exposed and any equipment likely to contain sensitive components (e.g., switchgear) is well vented. Therefore, there is no impact of hot gas layer on temperature sensitive equipment for the fire scenarios in this fire zone. TB-1 4401 31 47,524 I Although this fire zone contains multiple electrical panels that may contain temperature sensitive equipment, the large ceiling height and volume of the area prevents the gas layer height from descending to the equipment heights. Therefore, there is no impact of hot gas layer on temperature sensitive equipment for the fire scenarios in this fire zone.
h) Response provided by ULNRC-05851 dated April 17, 2012.
i) The ability of National Inst itute of Standards and Technology (NIST) Fire Dynamics Simulator (FDS) to determine smoke detector response using the Cleary correlation was not verified and validated in NUREG-1824. However, the ability of FDS to predict detector activation has been verified and validated by Combustion Science and Engineering (CSE), Inc. and NIST. The analysis of Fire Compartment C-21 (also applicable to Fire Compartment C-
: 22) has been evaluated against these studies and the model was determined to be appropriate for use in the analysis. Technical details demonstrating the applicability of the validation studies have been documented in Report No. R1984-001-002, " Verification and Validation of
 
Fire Modeling Tools and Approaches for Use in NFPA 805 and Fire PRA Applications".
The use of Fire Dynamics Simulator to predict detector activation has been added to Section 4.5.1.2 and Attachment J to the Transition Report as shown in Attachment 1 and Attachment J to this enclosure. Discussion of the use of Fire Dynamics Simulator within the validated range of NUREG-1824 is contained in the response to FM RAI 01-e.
j) The developers of PyroSim (Thunderhead Engineering) confirmed that PyroSim is verified to build the input file correctly. A multi-level process is used to do this, including testing during    to ULNRC-05876
 
Page 15 of 136 development and running example problems through the software to ensu re the correct input data is written and results obtained. Selected examples from NUREG-1824, "Verification & Validation of Selected Fire Models for Nuclear Power Plant Applications, Volume 7: Fire Dynamics Simulator," are used for some of these example problems to ensure the input is written correctly. In addition, PyroSim has been in use by h undreds of users since 2006 and any discrepancies identified by these users are addressed in subsequent releases of the software. Details documenting how PyroSim is verified to build the input file correctly have been added to Report No. R1984-001-002, "Verif ication and Validation of Fire Modeling Tools and Approaches for Use in NFPA 805 and Fire PRA Applicat ions," Revision 1. Information on the use of PyroSim to generate the input files has been added to Report No. R1984-001-001, "Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling," Revision 0.
to ULNRC-05876 Page 16 of 136 Fire Modeling RAI 02 NFPA 805, Section 2.7.3.5, "Uncertain ty Analysis," states: "An un certainty analysis shall be performed to provide reasonable assurance that the performance criteri a 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 c onsidered in the contex t of the application.
: a. Please explain in detail the uncertainty analyses for fire modeling that was performed. Please describe how the uncertainties of the input parameters (geometry, Heat Release Rate (HRR), Response Time Index (RTI), etc.) were determined and accounted for and substantiate the statement in Appendix J of the Transition Report that states, "... the predictions are deemed to be within the bounds of experimental uncertainty ..."
Additional Justific ation Needed.
The intent of this RAI was for the licensee to discuss all of the relevant model output parameters described in Table 3-1 of NUREG 1824 (Table 4-1 of NUREG 1934), not just these three. Expand the response to include and justify (as applicable) all the relevant model output parameters described in Table 3-1 of NUREG 1824.
: b. During the audit, the NRC staff reviewed EPM Report No. R1984-001-001, "Fire Dynamics Simulator Analysis R0." The staff noted that cable tray obstructions were omitted in the FDS fire modeling analysis for Fire Areas C-21 and C-22. 
 
In a typical fire risk assessment, there are completeness uncertainties in the risk contribution due to scenarios not explicitly modeled (e.g., smoke damage), model uncertainties in the assessment of those scenarios that are explicitly modeled (e.g., uncertain ties 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 s ource fuel). Please justify why cable tray obstructions could be omitted in the FDS fire modeling analysis for Fire Areas C-21 and C-22.
Additional Justification Needed 
 
Provide additional explanation of why hot gas and smoke movement is not affected. During the audit walkdown, NRC staff observed numerous cable trays and other obstructions directly above and adjacent to the ignition source but still within the beam pocket. The calculations assume three cable trays stacked vertically above the fire source location. Provide a detailed explanation of why the additional obstructions are not expected to break up and delay the development of the plume/HGL necessary to activate a detector and/or sprinkler in the beam pocket. In addition, provide a description of the decision process and criteria used for omitting specific obstructions from the FDS analyses.
 
The results appear to be counter-intuitive and it may be necessary to determine up to what extent the revised geometry in the FDS input file is representative of that in the plant. Provide the FDS input files and pictures that s how the obstructions in areas C-21 and C-22. to ULNRC-05876
 
Page 17 of 136 Response to Fire Modeling RAI-02 a) The following supersedes the response to this RAI provided by ULNRC-05851 dated April 17, 2012. Fire modeling in support of the transition has been performed within the Fire PRA, utilizing
 
codes and standards developed by industry and NRC staff which have been verified and validated in authoritative publications, such as NUREG 1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications," Final Report, dated April 2007. In general, the fire modeling in support of the Fire Risk Evaluations has been performed using conservative methods and input parameters that are based upon NUREG/CR- 6850, "Fire PRA Methodology for Nuclear Power Facilities," Final Report, September 2005. This pragmatic
 
approach is used given the current state of know ledge regarding the uncer tainties related to the application of the fire modeli ng tools and associated input parameters for specific plant configurations. A characterization of uncertainties associated with detailed fire modeling has been documented in Section 9 of each fire zone-specific Detailed Fire Modeling Report and is
 
summarized below:  The detailed fire modeling task develops a probabilistic output in the form of target failure probabilities and are subject to both aleatory and epistemic uncertainty. Appendix V of NUREG/CR-6850 sugge sts 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. Thes e distributions should be based on the variation of experimental results as well as the analyst's 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 sp ecific characteristics of the s ource and the configuration of secondary combustibles. Aleatory uncertainties identified within the fire modeling parameters include:
* Detector response reliab ility and av ailability
* Automatic suppression system re liability and availability
* Manual suppression reliability with respect to time available
 
Epistemic uncertainties which impact the zone of influence and time to damage range include:
* Heat release rates (peak HRR, time to reach peak, steady burning time, decay time)
* Number of cabinet cable bundles
* Ignition source fire diameter
* Room ventilation conditions
* Sprinkler Response Time Index (RTI), C factor, and activation temperature
* Detector activation temperature, geometry and obscuration activation    to ULNRC-05876
 
Page 18 of 136
* Soot yield
* Fire growth assumptions (cable tray empirical rule set, barrier delay)
* Cable fire spread characteristics for horizontal and vertical trays
* Transient fires (peak HRR, time to reach peak, location factor, detection time)
* Oil fires (spill assumptions)
* Assumed target location
* Target damage threshold criteria
* Manual detection time
* Mean prompt suppression rate
* Manual suppression rate
* Welding and cutting target damage set
* Transient target impacts With respect to the PRA, a quantitative char acterization has not been developed as the quantitative results are conservatively biased for key contributors. Ra ther 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 e nd. Due to the uncertainty with each of these parameters, the fire modeling task has se lected conservative values for each.
Fire models should be created with a substantial safety margin. Per NEI 04-02, there is no clear definition of an adequate safety margin. Howeve r, it should be sufficiently large so as to bound the uncertainty within a partic ular calculation or application. The detailed fire modeling calculations provide a list of items that are modeled conservati vely and that provide safety margin. Some examples include the following items:
* Fire scenarios involving electri cal cabinets (including the elec trical split fraction of pump fires) utilize the 98th percentile HRR for the severity factor calculated out to the nearest FPRA target. This is considered conservative.
* The fire elevation in most cases is at the top of the cabinet or pump body. This is considered conservative, since the combustion process will occur where the fuel mixes with oxygen, which is not always at th e top of the ignition source.
* The radiant fraction utilized is 0.4. This represents a 33% increase over the normally recommended value of 0.3.
* The convective heat release rate fraction utilized is 0.7. The normally recommended value is between 0.6 and 0.65, and thus the use of 0.7 is conservative.
* For transient fire impacts, a large bounding transient zone assumes all targets within its ZOI are affected by a fire. Time to damage is calculated based on the most severe (closest)
 
target. This is considered conservative, since a transient fire would actually have a much    to ULNRC-05876
 
Page 19 of 136 smaller zone of influence and varying damage times. This approach is implemented to minimize the multitude of transient scenarios to be analyzed.
* For hot gas layer calculations, no equipment or structural steel is credited as a heat sink, since the closed-form correlations used do not account for heat loss to these items.
* Not all cable trays are filled to capacity. By assuming full, this provides conservative estimates of the contribution of cable insulation to the fire and the corresponding time to damage.
* As the fire propagates to secondary combustibles, the fire is conservatively modeled as one single fire using the fire modeling closed-form correlations. The resulting plume temperature estimates used in this analysis are therefore also conservative, since in actuality, the fire would be distributed over a large surface area, and would be less severe at the target location.
* Target damage is assumed to occur when the exposure environment meets or exceeds the damage threshold. No additional time delay due to thermal response is given.
* The fire elevation for transient fires is 2-feet. This is considered conservative since most transient fires are expected to be below this height or even at floor level.
* Oil fires are analyzed as both unconfined and confined spills with 20-minute durations.
Unconfined spills result in larg e heat release rates, but us ually burn for seconds. The oil fires have been conservatively analyzed for 20-minutes to account for the uncertainty in the oil spill size.
* High energy arcing fault scenarios are conservatively assumed to be at peak fire intensity for 20-minutes from time zero, even though the initial arcing fault is expected to consume the contents of the cabinet and burn for only a few minutes.
* Fire brigade intervention is not credited prior to 85-minutes. Fire Brigade drills indicated that typical manual suppression times can be expected to be much less (i.e., 20 minutes).
All of the fire models used at Callaway Plant and listed in Attachment J of the Callaway Plant NFPA 805 Transition Report were evaluated for experimental uncertainty. The degree to which each model falls within or outside of experimental uncertainty is given in NUREG-1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications", Final Report, dated April 2007. Each model is discussed as follows: Hot Gas Layer Temperature using FDTs The predictive capability of these parameters using FDTs is characterized as YELLOW+ according to Table 3-1 of NUREG-1824. to ULNRC-05876
 
Page 20 of 136 A YELLOW+/- characterization is given: "If the first criterion is satisfied and the calculated relative differences are outside the experimental uncertainty but indicate a consistent pattern
 
of model over-prediction or under-predicti on, then the model pred ictive capability is characterized as YELLOW+ for over-predi ction, and YELLOW- for under-prediction. The model prediction for the specific attribute may be useful within the ranges of experiments in this study, and as described in Tables 2-4 and 2-5, but the users should use caution when
 
interpreting the results of the model. A complete understanding of model assumptions and scenario applicability to these V&V results is ne cessary. The model may be used if the grade is YELLOW+ when the user ensures that model over-prediction reflects conservatism. The user must exercise caution when using models with capabilities described as YELLOW+/-."
NUREG 1824, Volume 3, Section 6.1 states that: "The FDTs models for HGL temperature
 
capture the appropriate physics and are based on appropriate empirical data. FDTs generally over-predicts HGL temperature, outside of uncer tainty." The over-prediction is expected to lead to conservative results and increased safety margin.
Hot Gas Layer Height and Temperature using FDS The predictive capability of these parameters in FDS is characterized as GREEN according to Table 3-1 of NUREG-1824. A GREEN characterization is given: " If both criteria are satisfied (i.e., the model physics are appropriate for the calculation being made and the calculated relative differences are within or very near experimental uncertainty), then the V&V team concluded that the fire model prediction is accurate for the r anges of experiments in this study , and as described in Tables 2-4 and 2-5. A grade of GREEN indicates the model c an 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 w ithin the ranges of the parameters defined by the experiments, it is validated."
The NUREG-1824, Volume 7, Section 6.1 summary states:
"FDS is suitable for predicting HGL temperature and height, with no specific caveat s, 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 designa tion is discussed abov e under the "Hot Gas Layer Height and Temperature using FDS" h eading. Specifically, the GREEN designation was assigned to the CFAST HGL temperature parameter calculated in the fire compartment of origin. Compartments remote from the fire we re assigned a yellow desi gnation. Callaway Plant    to ULNRC-05876
 
Page 21 of 136 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-pre dicted and HGL height somewhat lower than experimental measurements. These differences are likely attributable to simplifications in the model dealing with mixing between the layers, entrainment in the fire plume, and flow through vents. Still, predictions are mostly within 10% to 20% of experimental measurements."
Ceiling Jet Temperature us ing Alpert Correlation The predictive capability of this parameter using the Alpert correlation in the fire model FIVE is characterized as YELLOW+ according to Table 3-1 of NUREG-1824. The YELLOW+
designation is discussed above under the "Hot Gas Layer Temperature using FDTs" heading. Specifically NUREG-1824, Volume 5, Section 6.2 summary states:
"The Alpert correlation under-predicts ceiling jet temperatures in comp artment fires with an es tablished hot gas layer. This result is expected because the correlation was developed without considering HGL effects.
The original version of FIVE accounted for HGL effects by adding th e ceiling jet and HGL temperature. This practice result s in consistent over-predictions of the ceiling jet temperature.
The approach of adding ceiling jet temperatures to th e calculated hot gas layer continues to be the recommended method for FIVE-Rev1 users. Based on the above discussion, a classification
 
of Yellow+ is recommended if HGL effects on the ceiling jet temperature are considered using the approach described in the above bullet. The Al pert correlation by itself is not intended to be used in rooms with an established hot gas layer."
The approach of adding the hot gas layer temperature to the ceiling jet temperature was not used for fire modeling at Callaway Plant. The primary application of the ceiling jet correlation at Callaway Plant was the determination of detection and suppression timing, in which the ceiling jet velocity is a sub-model in the analysis. Including the effects of a hot gas layer would result in shorter detection and suppression times, and therefore the use of the ceiling jet correlation at Callaway Plant is considered cons ervative. The use of the ceiling jet correlation for target damage is almost always bounded by th e use of the point source radiation model and is discussed in detail in the response to RAI 01-b. Plume Temperature using FDTs The predictive capability of this parameter using FDTs is characterized as YELLOW- according to Table 3-1 of NUREG-1824. The YELLOW- designation is discussed above under the "Hot Gas Layer Temperature using FDTs" heading. The NUREG-1824, Volume 3, Section 6.2 summary states: "The FDTs model for plume temperature is based on appropriate empirical data and is physically appropriate. FDTs generally under-predicts plume temperature, outsi de of uncertainty, becau se of the effects of the hot gas layer on test measurements of plume temperature. The FDTs model is not appropriate for predicting the plume temperatur es at elevations within a hot gas layer."    to ULNRC-05876
 
Page 22 of 136 The use of the FDTs plume correlation for fire modeling applications at Callaway Plant was used within the limitations given in NUREG-1824. The effects of a the plume and hot gas layer interaction were analyzed and documente d in detail in Appendix B of R1984-001-002, "Verification and Validation of Fire Modeling Tools and A pproaches for Use in NFPA 805 and Fire PRA Applications." The use of the FDTs plume correlation was used in accordance with the results of this analysis. Plume Temperature using FDS The predictive capability of this parameter using FDS is characteri zed 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 uncerta inty 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 sh ould 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 validat ion 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 appropria teness of the model prediction to the fire scenario for which it is being used."
The NUREG-1824, Volume 7, Section 6.3 summary states:
"The FDS hydrodynamic solver is well-suited for this application. FDS over-predicts the lower plume temperature in BE #2
 
because it over-predicts the flame height. FDS predicts the FM/SNL plume temperature to within experimental uncer tainty. The simulations of BE #2 and the FM/SNL series are the most time-consuming of all six test series, mainly because of the need for a fairly fine numerical grid near the plume. It is important that a user understand that co nsiderable computation time may be necessary to well-resolve temperatures within the fire plume. Even with a relatively fine
 
grid, it is still challenging to accurately predict plume temperatures, especially in the fire itself or just above the flame tip. There are only nine plume temperature measurements in the data set. A more definitive conclusi on about the accuracy of FDS in predicting plume temperature would require more experimental data."
Per the guidance given in NUREG-1934, a D*/x ratio of 5 to 10 produces favorable FDS results at moderate computational cost. This guidance was used for the two Callaway Plant FDS studies that analyzed plume temperatures. The first is the plume and hot gas layer interaction study found in Appendix B of R1984-001-002, "Verification and Validation of Fire Modeling Tools and Approaches for Use in N FPA 805 and Fire PRA Applications" and the second is an analysis of suppression timing in Compartment C-31 found in Appendix C31 of R1984-001-001, "Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling," Revision 0. The D*/x ratio for the critical mesh used in each study is 5.13 and 6.04, respectively, ensuring that the mesh is fine enough to analyze plume temperatures in each case. In addition, the plume temperatures within the flaming re gion are not the focal point of either study. to ULNRC-05876
 
Page 23 of 136 Flame Height using FDTs The predictive capability of this parameter using FDTs is characterized as GREEN according to Table 3-1 of NUREG-1824. The GREEN designa tion is discussed abov e under the "Hot Gas Layer Height and Temperature using FDS" heading. The NUREG-1824, Volume 5, Section 6.3 summary states:
"The FDTs model predicted flame heights consistent with visual test observations."
Smoke Concentration using CFAST The predictive capability of this parameter in CFAST is characterized as YELLOW according to Table 3-1 of NUREG-1824. The YELLOW designation is discussed above under the "Plume Temperature using FDS" heading. The NUREG-1824, Volume 5, Section 6.6 summary states:
"CFAST is capable of transporting smoke throughout a compartment, assuming that the production rate is known and its transport properties are comparable to gaseous exhaust products. CFAST typically over-predicts the smoke concentration in all of the BE #3 tests, with the exception of Test 17. Predicted concentrations for open-door test s are within experimental uncer tainties, but those for closed-door tests are far higher. No firm conclusions ca n be drawn from this single data set. The measurements in the closed-door experiments are inconsistent with basic conservation of mass arguments, or there is a fundamental change in the combustion process as the fire becomes oxygen-starved."
Smoke concentration was analyzed in Ca lculation 17671-010b, "Callaway NFPA 805 Fire PRA - Main Control Room Fire Analysis," Revision 1, which was used to determine the probability of Main Control Room evacuation at Callaway Plant following a fire scenario. The Main Control Room is fully enclosed and separated from adjacent areas by vertical control boards and wallboard partitions and was therefore modeled with closed doors. The over-prediction of smoke concentration for closed-door tests as indicated in NUREG-1824 is expected to result in conservative results for this analysis. The smoke production rates used in the model are known and were derived from Table 3-4.16 of the SFPE Handbook of Fire Protection Engineering, 4 th Edition. Transport properties of the smoke are expected to be comparable to gaseous exhaust products.
Oxygen Concentration using CFAST The predictive capability of this parameter in CFAST is characterized as GREEN according to Table 3-1 of NUREG-1824. The GREEN designati on is discussed above under the "Hot Gas Layer Height and Temperature using FDS" heading. to ULNRC-05876
 
Page 24 of 136 The NUREG-1824, Volume 5, section 6.5 summary states: "
CFAST uses a simple user-specified combustion chemistry scheme based on a prescribed pyrolysis rate and species yields
 
that is appropriate for the applic ations studied. CFAST predicts the major gas species close to experimental uncertainty."
Radiant Heat using FDTs The predictive capability of this parameter in FDTs is characterized as YELLOW according to Table 3-1 of NUREG-1824. The YELLOW designation is discussed above under the "Plume Temperature using FDS" heading. The NUREG-1824, Volume 3, Section 6.4 summary states:
"The FDTs point source radiation and solid flame radiation model in general are based on appropriate empirical data and is physically appropriate with c onsideration of the simplifying assumptions. The FDTs point source radiation and solid flame radiation model are not valid for elevati ons within a hot gas layer. FDTs predictions had no clear trend. Th e model under- and over-predicted, outside uncertainty. The point source radi ation model is intended for pr edicting radiation from flames in an unobstructed and smoke-clear pat h between flames and targets."
Only the FDTs point source radiation model was used for fire modeling at Callaway Plant.
NUREG- 1824 indicates that there is no clear trend in under or over-prediction for the point source model. The model over-predicted heat flux for locations immersed in a hot gas layer, which is likely due to smoke and the HGL preventing radiation from reaching the gauges. This over-prediction is expected to lead to conservative results and increased safety margin. In a smaller number of cases, the model under-predicted heat flux due to cont ribution of radiation from the HGL. In order to account for this potential under-prediction, conservatism has been built into the use of the radiation model at Callaway Plant, including the use of a radiant heat release rate fraction of 0.4.
 
In addition, NUREG-1824 indicates that the point source model is not intended to be used for locations relatively close to the fire. For fire modeling at Callaway Plant, targets located close to the fire have conservatively been faile d within the early stages of fire growth.
Radiant Heat using FDS The predictive capability of this parameter in FDS is characterized as YELLOW according to Table 3-1 of NUREG-1824. The YELLOW designation is discussed above under the "Plume Temperature using FDS" heading.
Even though the FDS Radiant Heat Mode l was given a Yellow designation, NUREG 1824, Volume 7, Section 6.8 states that:
"FDS has the appropriate radi ation and solid phase models for predicting the radiative and convective heat flux to targets, assuming the targets are relatively simple in shape. FDS is capable of predicting the surface temperature of a target, assuming that its shape is relatively simple and its composition fairly uniform. FDS predictions of heat flux and surface temperature are genera lly within exper imental uncertain ty, 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."
to ULNRC-05876
 
Page 25 of 136 FDS was used to calculate radiant heat exposur e at Callaway Plant for two applications. The first application was to determine the radiant he at 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 calcu lated 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 limita tions identified, it is concluded that the FDS predictions of heat flux is within experimental uncertainty.
b) Although the FDS model did not exac tly replicate the field conditions in terms of cable tray obstructions, a sensitivity study determined that omitting these does not significantly affect the output parameters being evaluated in the FDS model (i.e., automatic de tection and suppression system activation). The detection and suppression timing determined in the sensitivity study does not change the target damage set determined for the scenarios. The analysis is documented in Sections C21.7.2 and C22.7.2 of the FDS Report R1984-001-001, "Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling". The FDS computer input file is provide d in Enclosure 2.
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Page 26 of 136 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 discus sion was found in the Transition Report, with respect to how the input for the algebraic mode ls were established fo r 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 speci fic discussion in the Tr ansition 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 fi re modeling report R1984-001-001 "Fire Dynamic Simulator Analysis to Support Detailed Fire Mode ling," Rev. O. states in several places (all Appendices) that, "the mesh size reflects the finest mesh feasibly allowable with the given computer resources." Please explain why the mesh size used is within the validated range and confirm whether a grid sensitivity study was performed or justify why such a study was not performed.
Additional Documentation Required: 
 
The draft response states that a sensitivity study was conducted for Mesh 1 in Fire Area C-1 to show that results of a 0.05-m mesh would yield the same conclusions as the 0.1-m mesh used in the original analysis. Provide th e grid sensitivity study on the portal.
: d. During the audit,  the NRC staff noted that Section A11.1 of fire modeling report, R1984-001-001 discussed how the analysis performed for Fi re Area C-31 was applied to Fire Area A-11 since the room dimensions for both spaces are comparable. However, this discussion does not describe how the ignition source location and the radial distance between the fire source and the sprinkler was selected.
Please explain how the assumption to use the FDS an alysis 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 a nd 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 verifi cation and validation for estimating sprinkler activation times. However, the major inputs used in the determination of suppression    to ULNRC-05876
 
Page 27 of 136 (determination of gas temperatures) have been validated." Based on this statement, it was not clear to the staff how the sprinkler activation time was determined. Please explain how the sprinkler activation time was calculated in the FDS analysis.
 
Additional Documentation Required The Heskestad/Bill equation and part of the text appear to be missing from the non-docketed section. Provide the missing information in the response.
: f. During the audit, the NRC staff noted that different material properties were used in the FMDB analysis as in the FDS analysis for the same fire areas (A-11, C-21, etc.). For example. in Calculation No. KC-49, the materi al properties used in the FD S analysis for concrete is different from that used in the FMDB and transient datasheet analysis. The thermal conductivity and density in the FMDB are 1.6 Watts per meter Kelvin (W/m-K) and 2400 kilograms per cubic meter (kg/m
: 3) as opposed to 1.0 W/m-K and 2100 kg/m 3 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 expl ain what effect the difference in material properties used in the analyses has on the conclusions.
Additional Clarification Required
 
The licensee stated that the difference in materi al properties does not significantly affect the thermal inertia and therefore this difference will not affect the results of the analysis. Provide justification for the statement "- the difference in the thermal inertia valu es is not significant".
: g. During the audit, the NRC staff noted that fi re modeling report R 1984-001-001 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 igniti on source which would not activate sp rinklers prior to ignition of secondary combustibles.
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Page 28 of 136 Based on this discussion in the report, it was not clear to the NRC sta ff how these assumptions were verified. Please explain how the heat releas e rate profiles chosen were conservative for the purposes of damage assessment and sprink ler activation. In addi tion, please apply this response to the analysis conducted for the other two cable chase fire areas (C-30 and C-31) analyzed with FDS.
Additional Justification Needed The licensee's approach recognizes the fact that there is a trad e-off between choosing a low vs. a high heat release rate. The former delays dete ctor and sprinkler activa tion but also results in less damage. To be conservative a relatively high peak heat release rate (317 kW) and a relatively slow growth rate (8 min to peak heat release rate) were used. Explain why only these two values were used and why a different set of equally plausible values would not result in greater risk.
: i. During the audit, the NRC staff noted that fi re 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 FD S analysis and provide technical justification for the conclusion that the slice file output show s that fire location does not affect the sprinkler activation times.
Additional Documentation Required Part of the draft response between pages 1 and 2 appears to be missi ng. Provide the missing information in the response.
: j. During the audit, the NRC staff noted that Section C21.2 of fire modeling report R 1984-001-001 states, in part, that, "the purpose of the FDS simulation was to determine the time at which the ceiling-mounted quick-response sprinklers in this fire compartment would activate as a result of a transient fire." However, in the para graph that follows, it is stated that the sprinklers were given an RTI of 130 milliseconds0.5 (m-s0.5), which is a value more typical of a standard response sprinkler. Please state what type of sprinklers are in the lower Cable Spreading Room (CSR) and also provide a justification for the RTI used in the analysis.
Additional Justification Needed See action for RAI 3k response below.
: k. During the audit, the NRC staff noted that Section C21.3.5 of fire modeling report R1984-001-001 states that standard response sprinklers are used in the CSR and therefore an RTI of 130 (m-s)0.5 was used for the analysis. The licensee justified this value for the RTI by way of    to ULNRC-05876
 
Page 29 of 136 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 res ponse to the upper CSR (Fire Area C-22).
Additional Justification Needed
 
In the FDS analysis of fire area C-21, a sprinkler head RTI of 130 (ms)0.5 was used based on NUREG-1805. This value was justified because standard response sprinklers have an RTI of 80(m-s)1/2 or higher and the use of 130(m-s) 1/2 is therefore conservative. The draft response refers to a NIST study, which is cited as the basis of the default values in NUREG 1805. However, the objective of that study was to compare four sprinkler activation models. Another study by the same author and published in Operation of Fire Protection Systems (special addition to the NFPA Handbook), shows that the typical range for standard response sprinkler RTIs is approximately 100-350(m-s) 1/2. Provide justification for the use of an RTI of 130(m-s)1/2 in lieu of the more conservative values reported in the literature.
 
A typical range for standard re sponse sprinkler response time index (RTIs) is 100-350 (m-s) 1/2. Perform a sensitivity analysis to substantiate the use of an RTI of 130 (m-s) 1/2. I. During the audit, the NRC staff noted that Section C21.3.5.1 of fire modeling report R 1984-001-001 discusses why a 45 kilowatt (kW) initiating fire was considered more conservative than a 69 kW initiating fire, in terms of sp rinkler activation and ignition of secondary combustibles. It was not clear to the staff how this conservatism was ve rified. Please explain how heat release rate profiles chosen were conservative for the purposes of damage assessment and sprinkler activation.
In addition, please apply this res ponse 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 suppres sion 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 stat ed 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, pl ease apply this response to the analysis conducted for the other upper CSR (C-22) analyzed with FDS.
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Page 30 of 136
: o. During the audit, the NRC staff noted that Section C21.5 of fire modeling report R 1984-001-001 states that "The FDS analysis results for Fire Compartment C-22 ar e 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 wa s 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 analys is with the licensee. During this discussion, NRC was told that it was assumed that a fire originating in the Equipment Cabinet Area (ECA) was assumed to not be able to propagate into the MCR. Please provide a basis for this assumption.
 
Additional Information Needed The concern also included the possibility of HGL forming above the MCR cabinets, as they did not extend to the ceiling and also any other means of fire spread via the open ceiling area, not
 
just cabinet-to-cabinet directly. Exposed cables, if any, could also be targets along which fire could propagate between the two areas.. The draft response disc usses why direct cabinet-to-cabinet fire propagation is precl uded, but does not discuss other potential modes of fire spread.
Provide a discussion of other potential m odes of fire spread in the response.
: q. During the audit, the NRC staff reviewed Sc ientech 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 clarif y whether there is unqualified cable in the control room and if so, what is the ra tio of unqualified to qualified cable.
: r. During the audit, the NRC staff reviewed Sc ientech Calculation 17671-010b, and discussed the analysis with the licensee. In Table 1 of Attachment 1 (Control Room Evacuation Study) of Scientech Calculation 17671-010b, the modeled fire scenarios are provided. For single cabinet fires, both qualified and unqualified cabling was used in the calculation of evacuation times.
However, for the multicabinet fire scenarios, only qualified cable was considered in the calculation of evacuation times. Please explain why unqualified cable was not considered for multi-cabinet fires.
: s. During the audit, the NRC staff reviewed Scientech Calculation 17671-101b, as well as Attachment 1 (Control Room Evacuation Study).
The fuel combustion properties for qualified and unqualified cable are provided in this report. The heat of combustion (HOC) for qualified and unqualified cable is given as 28.3 and 20.9 me gajoules per kilogram (MJ/kg), respectively.
It is not expected that the HOC for an unqualifie d cable would be lower than a qualified cable. Please confirm these material values and also explain how the HOC mate rial property is used in the analysis. to ULNRC-05876
 
Page 31 of 136
: t. Please provide the FDS input file s for the detailed FDS fire modeling conducted as described in EPM document Nos. R1984-001-001 and R1984-001-C1, Detailed Fire Modeling Report -
FDS Analysis of HDPE Pipes (Draft B).
Additional Information Needed The Society of Fire Protection Engineers (SFPE) Handbook lists two se ts of values for the type of qualified cable that is present in the MCR (XLPE/XLPE according to the MCR abandonment study report). XLPE/XLPE cable #1 in the SFPE handbook table has a soot yield of 0.12 g/g and a heat of combustion of 28.3 kJ
/g. The values for cable #2 are 0.12 g/g and 12.5 kJ/g, respectively. Explain why the soot yield and heat of combus tion for cable #1 were used in the analysis.
 
Response to Fire Modeling RAI-03 a) The approach for fires involving multiple combustibles was to calculate the heat release rate of each individual fire as a function of time, and then use the combined total heat release rate as the input to the algebraic models. Conservative heat release rates were determined from NUREG/CR-6850, and the rules for propagation to cable trays, and fire spread rates all followed the FLASH-CAT model found in NUREG
/CR-7010. This approach is considered appropriate for the following reasons:
* The approach is endorsed in Secti on 3.2.2.2 of NUREG-1934, second draft report for comment, which discusses summing up i ndividual heat release rates for use in algebraic models:
"The heat release rate from the cable tray can be added to the heat release rate of
 
the cabinet to determine a combined heat release rate as a function of time. This total rate can then be used in the vari ous models as an approximation of the heat release rate as a function of time."
* Using the sum of all heat release rates is expected to result in conservative estimation of zone of influence as calcula ted by the algebraic models. In a realistic setting, each individual fire taken separately would create smaller zones of influence than that calculated for one large, combined fire. This is in part due to the expected interference of the base fire on the plume entrainment and flame heights of the secondary combustible fires, resulting in the reduction of the effective mass burning rate of the secondary combustible fires. In addition, the obstructing fires could create an environment where the fire would be oxygen limited.
* The fire diameter used as the input to the algebraic models is equal to the fire diameter of the original source fire and remains unchanged throughout the burning duration of the fire. In reality, a spreading fi re will have an increasingly larger fire    to ULNRC-05876
 
Page 32 of 136 diameter. The use of the source diameter is considered more severe for plume and flame height correlations, as the use of a small diameter results in a stronger plume and thus larger vertical zone of influence values.
* Burnout was considered; however, spread along cable trays was modeled until 85 minutes. NUREG/CR-7010, Secti on 9.2.4 states that the a ssumption that the fire will spread laterally until all cable is consumed is conservative, as this phenomenon was not observed in many of the multiple tray experiments. Assuming total consumption of all cables leads to conser vative heat release rates and zone of influence calculations.
b) The Callaway Plant Main Control Room (MCR) evacuation study used Consolidated Model of Fire Growth and Smoke Transport (CFAST) models as described in Attachment A to calculation 17671-10b. A summary of the plant drawings and plant procedures used as inputs are provided below. Fire Modeling RAI-01-f provides additional information about parameter selection and validation of usage. The CFAST model input files for the Callaway Plant MCR
 
are provided in Enclosure 2.
* Control Building HVAC System design basis description document, ULDBD-GK-001, Rev. 0
* Control Room Inaccessibility procedure, OTO-ZZ-00001, Rev. 32
* Computer Room & Control Room Detailed Floor Plans, EL 2047'-6", Drawing A-2337 Rev. 11
* Heating, Ventilation & Ai r Cond. Control Building EL. 2047'-6", Drawing M-2H3611 Rev. 7
* Heating, Ventilation & Air Cond. Control Building Secti ons & Details, Drawing M-2H3901 Rev. 6
* Hot Lab, Counting Room, Control Room & Co mputer Room Reflected Ceiling Plans, Drawing A-2332 Rev. 9 c) NUREG-1824, "Verification and Valid ation of Selected Fire Mode ls for Nuclear Power Plant Applications," Final Report, April 2007, defines the validated range as a D*/x value from 4 to 16. Further, NUREG-1934, "Nuclear Power Plant Fire Modeling Application Guide," Second Draft for Comment, July 2011, states that D*/x  ratios of 5 to 10 usually produce favorable results at a moderate computational cost. Finer meshes would not necessarily provide more accurate results while causing a dramatic increase in computation time. 
 
For each FDS analysis documented in R1984-001-001, "Fire Dynamics Simulator (FDS)
Analysis to Support Detailed Fire M odeling," Revision 0, the D* and D*/x values were calculated. The mesh sizes utilized in the models were determined to be either acceptably refined based on industry guidance or a sensitivity study was performed to determine that the results and conclusions were valid. The statement "the mesh size reflects the finest mesh size feasibly allowable with the given computer resources" has been removed and the results of the D* analysis have been added to S ections A11.4.1, C31.4.1, C30.4.1, C21.4.1 and C22.4.1 of    to ULNRC-05876
 
Page 33 of 136 R1984-001-001 "Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling," Revision 0.
 
d) Response provided by ULNRC-05851 dated April 17, 2012.
 
e) Sprinkler activation times were predicted by the NIST Fire Dynamic Simulator (FDS) which uses the differential equation of Heskestad and Bill (published in Fire Safety Journal volume 14, 1988) to determine the link temperature. This method calculates sprinkler activation time based on the gas temperature and the sprinkler parameters. The sprinkler activation time is, therefore, reliant on the ability of FDS to predict ceiling jet and gas layer temperatures. NUREG-1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications," Final Re port, April 2007, assigned FDS a Green ranking for its ability to predict both ceiling jet and gas layer temperatures.
Based on the validation of the FDS model's ability to predict ceiling jet/gas temperature, and that the Heskestad and Bill method is documented in an authoritative publication, the use of FDS to determine sprinkler activation time is considered acceptable. This information has incorporated into Appendix H of Report R1984-001-002, "Verification and Validation of Fire Modeling Tools and Approaches for Use in N FPA 805 and Fire PRA A pplications," Revision
: 1.
f) The Fire Dynamics Tools (FDTs) presented in NUREG 1805, "Fire Dynamics Tools: Quantitative Fire Hazard Analysis Methods for the US Nuclear Regulatory Commission Fire Protection Inspection Program," December 2004, suggest thermal properties for common boundary materials. The Fire Dynamics Simulator (FDS) developed by the National Institute of Standards and Technology (NIST), comes equipped with a material properties database that also contains the thermal properties for a rang e of common materials. The Callaway Plant fire models used the thermal propertie s as suggested by the specific tool as the values for the installed boundary materials.
 
The Fire Modeling Database (FMDB) utili zes the equations provided by the NUREG 1805 FDTs and the FMDB results have been verified and validated against those generated by the NUREG 1805 spreadsheets. Therefore, the calculations performed using the FMDB used the suggested material properties provided in NUREG 1805 FDTs 2.1, 2.2 and 2.3.
 
The material properties used in the NIST FDS models are taken from the material properties database provided by NIST for use in FDS. The ma terial properties in the database have been reviewed against the materi al properties found in the SF PE Handbook of Fire Protection Engineering, 4 th Edition, and match the SFPE suggested values.
 
The NUREG 1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications," Final Report, April 2007, FDS verification and va lidation study includes the following discussion on the selection of material properties:
Some of the property data needed by FDS are commonly available in fire protection engineering and materials handbooks. Depending on the application, properties for specific    to ULNRC-05876
 
Page 34 of 136 materials may not be readily available (especially burning behavior at different heat fluxes). A small file distributed with the FDS software co ntains a database with thermal properties of common materials. This data are given as exa mples, and users should verify the accuracy and appropriateness of the data.
The inputs used in the FDS models described in R1984-001-001, "Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling," Revision 0 employed this methodology by selecting the material properties from the FDS database and validating the parameters against the suggested values in the SFPE Handbook, 4th edition. Therefore, the use of these values is considered appropriate.
 
g) Response provided by ULNRC-05851 dated April 17, 2012.
 
h) Sprinkler activation and time to target damage are dependent on the heat release rate (HRR) of the fire and the temperatures generated at the sprinkler/target location.
A larger heat release rate will result in quicker, more severe target damage, but will also prompt suppression to activate earlier in the scenario. Consequently, a conservative heat release rate for target damage is a non-conservative heat release rate for sprinkler activ ation. Therefore, in order to avoid non-conservative assumptions for target damage or suppression activation, a conservative heat release rate was employed for both analyses. A detailed discussion is included in R1984-001-001, "Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling," Revision 0.
i) The slice files shown in R1984-001-001, "Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling," Revision 0 were in cluded in the report to demonstrate that the gas temperatures at the ceiling are relatively uniform. Although only th e slice file at the time of activation is included in the report, ceiling temperatures were observed to be uniform throughout the Smokeview simulation. Based on th e even heating of the upper gas layer, the model is not sensitive to minor variations in the sprinkler location with respect to the fire location selected. Since a boundi ng fire size and location were chosen and the gas layer is expected to be of uniform temperature, all sprinkler locations with respect to the fire location have been bounded by the analysis. The disc ussion in R1984-001-001 has been clarified to provide additional discussion of the use of the slice file.
j) Contrary to Section C21.2 of fire modeling report R1984-001-001, "Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling," Revision 0, the cable spreading rooms are not equipped with quick-response sprinklers. The installed ceiling-mounted sprinklers are standard response type. Fire Modeling Report R1984-001-001 has been re vised to correct the error. The sprinkler head in use in fire areas C-21 a nd C-22 is a Star Model E, 165 °F, Spray Nozzle that uses a fusible link. Star Sprinkler Co. is no longer in business and the specific Response Time Index (RTI) for the Model E is unknown. Sta ndard response sprinklers have a RTI value of 80 m*s or more as defined by NFPA 13 200 2 edition. Because the specific RTI is unknown the NUREG-1805, "Fire Dynamics Tools: Quantitative Fire Hazard Analysis    to ULNRC-05876
 
Page 35 of 136 Methods for the US Nuclear Regulatory Commission Fire Protection Inspection Program," December 2004 FDT Section 10.3.3 generic RTI of 130m*s for standard response sprinklers with a thermal link was utilized. The RTI is a laboratory test obtained value that rates the thermal sensitivity of a sprinkler and the RTI is used primarily to categorize a sprinkler head
 
as either Quick Response or Standard Response. In field installed conditions numerous factors affect actual sprinkler activation times including sprinkler temp erature ratings, ceiling height, fire distance to the sprinkler, fire heat release rates, and sprinkler head distance from the ceiling. For C-21 and C-22 sprinkler activation is assumed after three trays are damaged or the equivalent of 14 minutes when following the EPRI cable tray fire propagation guidance. This sprinkler activation time is realistic and acceptable for use in the FPRA analysis.
 
k) Contrary to Section C21.3.5 of fire m odeling report R1984-001-001, the cable spreading rooms are not equipped with quick-response sprinklers. The installed ceiling-mounted sprinklers are standard response type. Fire Modeling Report R1984-001-001 has been revised to correct the error.
 
The sprinkler heads used in fire areas C-21 and C-22 are a Star Model E, 165 degree F, spray nozzle that uses a fusible link. Star Sprinkler Company is no longer in business and the specific Response Time Index (RTI) for the Model E spray nozzle is unknown. Standard response sprinklers have a Response Time Index (RTI) value of 80 (m-s) 1/2 or more as defined by NFPA 13, 2002 edition. Because the sp ecific RTI is unknown, the NUREG-1805 FDT Section 10.3.3 generic RTI of 130 (m-s) 1/2 for standard response sprinklers with a thermal link was utilized. The RTI value is obtained from laboratory testing which rates the thermal sensitivity of a sprinkler. The RTI is used primarily to categorize a sprinkler head as either Quick Response or Standard Response. In field installed conditions numerous factors affect actual sprinkler activation times including sprinkler temperature ratings , ceiling height, fire distance to the sprinkler, fire h eat release rates, and sprinkler head distance from the ceiling. The FDS analysis for C-21 and C-22 was performed only to determine the time to suppression activation. Fire propagation in the cable trays was determined using the timing of NUREG/CR-6850 Appendix R. Based on this guidan ce, cable tray ignition occurs as shown
 
below:
Cable Tray Ignition Time (minutes)
Tray #1 5 Tray #2 9 Tray #3 12 Tray #4 14 Tray #5 15 Tray #6 16 The FDS analysis determined that sprinkler activation will occur prior to ignition of cable tray #4, which occurs at 14 minutes.
 
The detailed fire modeling task develops a probabilistic output in the form of target failure
 
probabilities that are subject to both aleatory and epistemi c uncertainty. Appendix V of    to ULNRC-05876
 
Page 36 of 136 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 distri butions should be based on the variation of experimental results as well as the analyst's judgment. In addition, to the extent possible more than one fire model can be applied and probabilities assigned to the outcome which describe the degree of belief that each model is the correct one.
 
The propagation of fire is described by a fire model (represented by a fire growth event tree), which addresses the specific characteristics of the source and the configuration of secondary combustibles. 
 
Aleatory uncertainties identified within the fire modeling parameters include:
* Detector response reliab ility and av ailability
* Automatic suppression system re liability and availability
* Manual suppression reliability with respect to time available 
 
Epistemic uncertainties which impact the zone of influence and time to damage range include:
* Heat release rates (peak HRR, time to reach peak, steady burning time, decay time)
* Ignition source fire diameter
* Room ventilation conditions
* Fire growth assumptions (cable tray empirical rule set, barrier delay)
* Cable fire spread characteristics for horizontal and vertical trays
* Transient fires (peak HRR, time to reach peak, location factor, detection time)
* Assumed target location
* Target damage threshold criteria
* Manual detection time
* Mean prompt suppression rate
* Manual suppression rate
* Transient target impacts
 
Due to the uncertainty with each of these parameters, the fire modeling has selected conservative values for each as discussed below.
* Transient fires were postulated to occur throughout the compartment even though, based on the configuration, not all areas are accessible or realistic locations for transient ignition sources.
* For transient fire impacts, a large bounding transient zone assumes all targets within its footprint and within the zone of influence are affected by fire. This is considered conservative, since a transient fire would actually have a much smaller zone of influence.
* The HRR of the initial transient fire used was 45 kW verse 69 kW. to ULNRC-05876
 
Page 37 of 136
* The transient fire location was assumed to be 2 ft. below the first tray. However, in actual plant configurations most trays are located more than 2 ft. above the assumed fire elevation.
* The radiant fraction is assumed to be 0.4. This represents a 33% safety margin over the normally recommended value of 0.3. In addition, the convec tive heat release rate fraction utilized is 0.7. The norma lly recommended value is between 0.6 and 0.65.
* No credit is given in the FDS model for metal cable tray bottoms or covers that would delay damage or fire growth. Some cable trays have tray bottoms and covers installed.
* The HRR used in the FDS model does not include cable tray risers. Where cable trays risers exist that are adjacent to horizontal trays in the damage assessment they are assumed damaged however for the HRR used in FDS they are not included because they significantly increased HRR and reduced sprinkler activation time.
* The transient fire location was chosen based on a worst case (farthest) distance from the sprinklers, in order to ensure th e most conservative delayed suppression activation was modeled.
* Target damage is assumed to occur when the exposure environment meets or exceeds the damage threshold.
* When suppression fails, whole room damage is always assumed to occur. This is considered conservative as the damage set and time to damage would be less severe.
* Fire brigade intervention is not credited prior to 85-minutes. Fire Brigade drills indicated that typical manual suppression times can be expected to be much less.
A sensitivity analysis was performed to determine the sprinkler activation times assuming only
 
the RTI value is changed over the range of RTI values allowed for a standard response sprinkler which is 80 to 350 (m-s) 1/2. As previously noted, nume rous factors affect actual sprinkler activation times including sprinkler temperature ratings, ceiling height, fire distance to the sprinkler, fire heat release rates, and  sprinkler head distance from the ceiling.
 
to ULNRC-Sin c pre d gen e sim i was stan d Pro t hea d port i s)1/2 i gen e ran g tim e -05876 RTI 80 130 165 225 350 c e Star Spri n dates industr y eral range o f i lar design s p discussed w d ard FM 20 0 ection". Th e d design util i i on of the al i s unlikely f o eral range o f g e of 130-22 e assumed i n ACT I TIM E 12 m i 13 m i 13 m i 14 m i 15 m i n kler is no l o y-wide use o f RTI values p rinklers. T h w ith a repres e 0 0, "Appro v e representa t i zes a fusibl elowable ran g or the sprin k f RTI for th e 5 (m-s)1/2  ba n the fire mo d Pa g IVATION E i n 33 sec i n 34 sec i n 52 sec i n 17 sec i n 28 sec nger in busi n of the RTI t efor the spri n he range of t e ntative fro m v al Standard tive from Fa c e plug whic h ge (80-120) k ler. Theref o e Star Model a sed on its d e deling for fi r g e 38 of 136 D E A C- 8 7- 2 5- 0 7+ 1+ 8 n ess and th e est, a specifi c nkler may b e t ypical RTI m Factory Mfor Auto m a t ctory Mutu a h would not (m-s)1/2. It w o re, based o n E sprinkler e sign. As s h r e areas C-2 ELTA TO A CTIVATIO N 7 sec 5 sec (-3%)
7 sec (< -1
%8 sec (2%)
8 9 sec e design of t h c RTI value e establishe d values for a Mutual. Fact o a tic Control M al indicated t be expecte d was also not e n feedback f r would reas o h own in the t 1 and C-22 f ASSUMED N %)    h e Star Mo d is not avail a d based on e xstandard re s ory Mutual i s M ode Sprin k t hat the Star d to fall with i e d an RTI v a rom Factor y o nably be e x t able, the sp rfall within t h d el E sprinkl e a ble. Howe v x perience w s ponse spri n s responsibl e klers for Fir e Model E sp r i n the lowes a lue over 22 y Mutual, th e x pected to f a rinkler acti v he range of t e r v er, the w ith n kler e for the e r inkler t 5 (m-e a ll in the v ation t he    to ULNRC-05876
 
Page 39 of 136 expected RTI values. The epistemic uncertainty due to any one input such as the sprinkler RTI is addressed by the fire modeli ng selecting conservative values for a number of the inputs to account for that uncertainty. The sprinkler RTI of 130 (m-s) 1/2  is within the expected range of RTI for a standard response sprinkler with th e Star Model E design based on discussions with an industry expert. The use of a sprinkler RTI value of 130 (m-s) 1/2  results in a sprinkler activation time that is realistic and acceptable for use in the FPR A analysis. Uncertainty in the sprinkler RTI value is addressed similar to other epistemic uncertainty by use of conservative parameters in other fire model inputs.
 
l) Sprinkler activation and time to target damage are mainly dependent on the heat release rate (HRR) of the fire and the temperatures generated at the sprinkler/target location. A larger heat release rate will result in quicker, more severe target damage, but will also prompt suppression to activate earlier in the scenar io. Consequently, a conservative heat release rate for target damage is a non-conservative heat release rate for sprinkler activation.
In order to avoid non-conservative assumptions for target damage or suppression activation, a di fferent heat release rate was employed for both analyses.
 
The cable spreading rooms contain a large number of horizontal cable tray stacks. The timing
 
of fire propagation between cable tray stack s is prescribed by NUREG/CR-6850, "Fire PRA Methodology for Nuclear Power Facilities,"
Final Report, September 2005 and the analysis follows this timing. Therefore, a fire that is quickly suppressed will be prevented from spreading into multiple trays and reduce the number of targets damaged by the spreading fire
 
and expanding zone of influence. The fire suppression analysis in R1984-001-001, "Fire Dynamics Simulator (FDS) Analysis to Suppor t Detailed Fire Mode ling," Revision 0 was constructed to ensure a conservative estimate of suppression activation. Th is critical fire size was determined to be 45kW, as cable trays are located 2-ft above the floor. A fire generating less than 45kW will impact only floor-based targets and will not cause ignition of the tray stacks, and would, therefore, be non-conservative with respect to damage.
 
However, selecting a fire larger than 45kW will result in increased gas layer temperatures. As Fire Dynamics Simulator (FDS) uses the temp erature of the gas layer to determine the link/bulb temperature of the sprinkler to predict activation, incr easing the HRR and gas temperature will cause suppression to activate earlier in the scenario. Therefore, in order to maximize fire growth and spread without prom oting suppression, the fire suppression analysis in R1984-001-001 used a 45kW fire instead of a 69kW fire.
 
The fire suppression analysis in R1984-001-001 is used only to determine the time to suppression activation given the prescribed fire scenario. Th e fire growth analysis is performed in the detailed fire modeling calcu lations KC-68, "Detailed Fire Modeling Report for Fire Compartment C-21" and KC-69, "Detailed Fire Modeling Report for Fire Compartment C-22."
 
The zone of influence for target damage was determined using a 69kW transient fire. The use of the larger initiating fire results in a conservative target damage set. Th e fire growth analyses    to ULNRC-05876
 
Page 40 of 136 and target damage sets are documented in De tailed Fire Modeling Calculations KC-68 and KC-69. m) The statement "this timing corresponds to ignition of the third cable tray in a stack" is meant
 
to describe that suppression activates after the third cable try ignites, based on the combined heat release rate of the three trays. Th e FDS analysis contained in R1984-001-001, "Fire Dynamics Simulator (FDS) Analysis to Suppor t Detailed Fire Mode ling," Revision 0 was performed to determine the time to suppressi on activation given the fire growth analysis developed in Calculation KC-68, "Detailed Fire Modeling Report for Fire Compartment C-21". Timing of cable tray ignition was determined using the empirical rule-set prescribed in NUREG/CR-6850, "Fire PRA Methodology fo r Nuclear Power Facilities,"
Final Report, September 2005 Appendix R.4.2. Since cable tray ignition time is not determined by the FDS analysis, the statement relating cable tray ignition time to suppression timing is not appropriate for inclusion in R1984-001-001 and has been removed.
 
n) The statement "the modeled configuration of a tr ansient fire in C-21 (a nd applicable to C-22) does not result in the formation of a hot gas layer before automatic suppression is actuated" has been removed from sections C21.5 and C22.5 of Report R1984-001-001, "Fire Dynamic Simulator (FDS) Analysis to Support Detailed Fire Modeling," Revision 0. The FDS analysis was performed to evaluate suppression activation time only and not to evaluate hot gas layer formation within the compartment. Therefore, this statement is not required in the report.
o) Response provided by ULNRC-05851 dated April 17, 2012.
p) Propagation of fire from the back panels in the electrical cabinet areas (ECA) to the main control board (MCB)was not considered due to th e configuration of the panels in the ECA.
Callaway Plant drawing J-24001 shows the configur ation of the ECA. The back wall of the main control board (RL028 to RL012) is a solid st eel plate. It has no fi re protection rating, but it also has no openings. The distance between the RP053 cabinets and the back of the main control board is approximately 5 feet. That is the same distance between the SB32, SB29 cabinets and the main control boards RL018-RL012. The distance from the single cabinet RP068 to the main control board is about 3 feet, but RP068 is a single bundle cabinet.
 
Appendix S.1 of NUREG/CR-6850 states cabinet to cabinet propagation can be ignored if cabinets are separated by a doubl e wall with an air gap. This is the case for the ECA back panels and the back of the main control board. This supports the engineering judgment to not consider propagation from the back panels in the ECA to the main control boards.
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Page 41 of 136 Other Fire Spread Mechanisms:
 
Horizontal Propagation via cable trays: The ECA has a ceiling height of 26 ft. Heat and combustion products from panel fires in the ECA will most likely rise straight up to the ceiling. Some of the panels in the ECA have cable trays exiting the top. All but 3 of the cable trays that exit the panels rise directly up to the upper cable spreading room  (which is above the Main Control Room(MCR)). For panels with vertical cable trays exiting the t op, fire propagation in the horizontal direction to the main control bo ard is highly unlikely. Th ere are 3 trays that originate from panel RK045E and cross over to the MCA (main control area) over the top of the MCB panels, over the MCR acoustical ceiling. The RK045E panels are more than 20 ft. from the main control board. Drawings show the bottom of the cable trays from RK045E are 9 ft. over the cabinets they pass over. The cables are IEEE rated so growth is postulated to be
 
very slow. Horizontal propagation via cable trays was dism issed based on engineering judgment given the geometry and material properties.
 
Hot gas layer: The probability of a hot gas layer from the ECA causing damage to the main control board is bounded by the probability that a hot gas layer causes evacuation of the MCR. The CFAST calculation provides scenarios for HGL formation in the MCA from fires in the ECA. ECA fires causing damage to MCB a nd NOT causing evacuation are not seen as a separate scenario.
 
q) All cable in the MCR cabinets is qualified to IEEE-383 standa rds. The MCR evacuation study evaluated qualified and u nqualified cable fires as part of a project requirement. At the time the MCR evacuation study was performed in 2009, the cable content of the MCR had not been verified. The scope of the evacuation study was set to be applicable to all eventualities. The runs for unqualified cable are not used in the analysis.
 
r) All cable in the MCR cabinets is qualified to IEEE-383 standards. When the MCR evacuation study was first performed in 2009, the cable cont ent of the MCR had not been verified. The scope of the evacuation study was set to be applicable to all eventualities. The MCR evacuation study evaluated qualified and unqualif ied cable fires as part of a project requirement. In 2009, only single cabin et scenarios were considered.
The MCR evacuation study was updated in 2011, to account for multi-cabinet fires. By that time, it had been verified that all cable was qualified, so there was no need to perform a multi
 
cabinet run with unqualified cable.
 
s) The heat of combustion for qualified and unqualified table was derived from table 3-4.16 of
 
the SFPE Handbook of Fire Protection Engineering, 4 th edition. They are different from what one would expect intuitively. The Callaway Plan t Fire PRA model is based on qualified cable throughout the plant (with the exception of some areas of th e Turbine Building) and is documented as such in the MCR calculation (1 7671-10b). As a sensitiv ity case, the CFAST model was also quantified with unqualified ca ble so both HOC values are shown in the documentation.
 
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Page 42 of 136 t) The FDS input files are provided in Enclosure 2. The FDS input f iles have also been added as Appendix I to Report No. R1984-001-001, "Fire Dynamics Simulator (FDS) Analysis to Support Detailed Fire Modeling,"
 
The SFPE Handbook Table 3-4.16 gives two sets of data for XPLE/XPLE cables. The first of which gives the HoC as 28.3 MJ/kg which is the valu e used in the analysis. The second set of data gives the HoC as 12.5 MJ/kg. The soot yield is given as 0.12 g/g in both cases. CFAST gives higher values of optical density with higher HoC values, and therefore shorter times to reach abandonment conditions and higher abandonment probabilities. The higher value for HoC is therefore conservative with respect to the time to abandonment based on optical
 
density. 
 
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Page 43 of 136 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 wi th clarification." The complia nce 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 clarifica tion is and whether or not the requi red actions are necessary for the entire chapter 3 attribute. Please clarify the use of the two-part compliance statement and what the clarification is intended to be.
Response to Fire Protection Engineering RAI-01 Response provided by ULNRC-05851 dated April 17, 2012.
Fire Protection Engineering RAI 02  In Attachment A of the LAR, Table B-1, on page A-33, the compliance statement for NFPA 805 Section 3.4.1(a)(1) states "complies with clarification." The compliance ba sis states "the industrial fire brigade complies with NFPA 600-2000 Edition. Excepti ons requiring further action are identified below."  It is unclear what the clarifica tion is and whether or not the requi red actions are necessary for the entire chapter 3 attribute. Please clarify the use of the two-part compliance statement and what the clarification is intended to be. 
 
Response to Fire Protection Engineering RAI-02 Response provided by ULNRC-05851 dated April 17, 2012.
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Page 44 of 136 Fire Protection Engineering RAI 03  In Attachment A of the LAR, Table B-1, on page A-37, the compliance statement for NFPA 805 Section 3.4.2 states "complies," however, implementation items are listed below. Please clarify whether "complies" is the correct compliance statement with the requirements in this section or if the plant complies with required action or both.
 
Response to Fire Protection Engineering RAI-03 Response provided by ULNRC-05851 dated April 17, 2012.
Fire Protection Engineering RAI 04  In Attachment A of the LAR, Table 8-1, on page A-39, the compliance statements for NFPA 805 Sections 3.4.2.3 and 3.4.2.4 state "comp lies, with required action," and the compliance basis states "see implementation item identified below." It was noted that there are no implementation items identified below these two sections. Please identify the required actions.
Response to Fire Protection Engineering RAI-04 Response provided by ULNRC-05851 dated April 17, 2012.
Fire Protection Engineering RAI 05  In Attachment A of the LAR, Table B-1, on page A-45, the compliance statement for NFPA 805 Section 3.4.4 states "complies with clarification." The compliance basis states "Equipment is provided for the fire brigade as required. Per visual inspection of equipment, it is in accordance with applicable NFPA codes, as documented in CAR 200902315." However the clarification is not apparent. Please identify the clarification used to support the compliance statement.
Response to Fire Protection Engineering RAI-05 Response provided by ULNRC-05851 dated April 17, 2012.
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Page 45 of 136 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 st ates 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 equipmen t shall be equivalent to the equipment supplied by three hose houses. The compliance basis states that equipment on two mobile units is provided, but does not specify the amount of equipment provided. Please clarify the actual equipment equivalency for the mobile units.
 
Response to Fire Protection Engineering RAI-06 Response provided by ULNRC-05851 dated April 17, 2012.
Fire Protection Engineering RAI 07 In Attachment A of the LAR, Table B-1, on page A-64, the compliance statement for NFPA 805 Section 3.6.2 states "complies with clarification." However the clarification is not apparent. Please identify the clarification used to support the compliance statement.
Response to Fire Protection Engineering RAI-07 Response provided by ULNRC-05851 dated April 17, 2012.
Fire Protection Engineering RAI 08 In Attachment A of the LAR, Table B-1, on page A-66, the compliance statement for NFPA 805 Section 3.6.4 states "compliance by previous NRC approval." The compliance basis for this element does not address the provision of this section to provide manual fire suppression in areas containing systems and components needed to perform nuclear safety functions following a safe shutdown earthquake. Although not addressed in the LAR, 10 CFR 50.48(c)(vi) states NRC requirements for licensees that wish to apply the exception to Section 3.6.4. Please describe how compliance is achieved with the requirement to provide manual fire suppression to protect nucle ar safety functions in the event of a safe shutdown earthquake.
Response to Fire Protection Engineering RAI-08 Based on NRC feedback, the compliance statement for this section is acceptable as stated and no further action is required.
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Page 46 of 136 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 KC0 08, which is located in th e control room. Please provide further discussion on these cl arifications, including a descript ion of the alarm process and how the alarming condition is communi cated to the operator(s).
Response to Fire Protection Engineering RAI-09 Response provided by ULNRC-05851 dated April 17, 2012.
Fire Protection Engineering RAI 10 On December 21, 2011, there was a fire in the B em ergency 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. Subseq uently, 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 e ffects this incident, if any, and any subsequent actions taken as a result of this incident, have on the NFPA 805 LAR and the transition process.
Response to Fire Protection Engineering RAI 10 Response provided by ULNRC-05851 dated April 17, 2012.
Fire Protection Engineering RAI 11 Section 4.1.2.3 and Attachment L, Approval Request 1, of the LAR describe th e 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 administrati ve and/or operating procedures used to ensure that the minimum required fire protection water supply remains available.
Response to Fire Protection Engineering RAI-11 Response provided by ULNRC-05851 dated April 17, 2012.
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Page 47 of 136 Fire Protection Engineering RAI 12 Table B-1, Criteria 3.5.1(b), Fire Flow Rate of the LAR indicates that compliance with this item is not applicable. However, in Approval Request 1, compliance to this requirement, namely, the 500 gallons per minute (gpm) hose stream requirement, is the basis for the request. Please reconcile the discrepancy. 
 
Response to Fire Protection Engineering RAI-12 Response provided by ULNRC-05851 dated April 17, 2012.
Fire Protection Engineering RAI 13 In Attachment A of the LAR, Table B-1, on page A-91, the compliance basis for NFPA 805 Section 3.10.9 does not provide adequate detail to conclude that the possibility of secondary thermal shock damage was considered for the design of the gaseous fire suppression systems at Callaway plant. Please provide additional information to justify the conclusion that Halon 1301 does not present a risk of secondary thermal shock.
Response to Fire Protection Engineering RAI-13 Additional details have been added to the License Amendment Request (LAR), Transition Report,  Table B-1, Section 3.10.9 compliance statement which describe that the Halon systems do not present a risk of secondary thermal shock. The revised Transition Report, Ta ble B-1, Section 3.10.9 is provided in Attachment A to this enclosure.
Fire Protection Engineering RAI 14 NFPA 805, Section 3.9.1 requires that water-based fire suppression systems be installed in accordance with the appropriate NFPA standa rd. During the audit, it was observe d 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 h eads 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 re sponse sprinklers in cable chases A-11, C-30, and C-31 has been completed. Please provide the basis and justification for compliance to the appropriate NFPA standard.
Response to Fire Protection Engineering RAI-14 Response provided by ULNRC-05851 dated April 17, 2012.
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Page 48 of 136 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 System s" 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 pressure s at the hose connections and whet her 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). Incl ude any prior approvals and any justification for meeting any other NFPA 14 requirements, as necessary. Please update the code conformance review calculation document as necessary.
Response to Fire Protection Engineering RAI-15 Response provided by ULNRC-05851 dated April 17, 2012.
 
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Page 49 of 136 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 comp ly with Frequently Asked Question (FAQ) 10-0059. The NRC staff noted that the information provi ded 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 ar e already included with in the scope of the Maintenance Rule program.
: b. A description of the process th at 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.
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Page 50 of 136
: c. A demonstration of how the monitoring program will address response to programmatic or training elements that fail to meet performance goals (examples include fire brigade response or performance standards and disc repancies 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 gui dance 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, industr y wide operating experience including whether this process will include both internal and external assessments and the frequency at which these assessments will be performed.
Response to Monitoring Program RAI-01 The LAR Transition Report Section 4.6.2 has been revised to align with the approved FAQ 10-0059 and its related closure memo. The revised Section 4.6.2 describes the process by which systems, structures, and components (SSCs) will be identi fied for inclusion in the NFPA 805 monitoring program, including the approach to be applied to a ny fire protection SSCs that are already included within the scope of the Maintenance Rule program. The revised LAR Transition Report Section 4.6.2 is provided in attachment 1 to this enclosure. Additionally, LAR Transition Report Attachment H and Attachment S have been updated to reflect that FAQ 10-0059 has now been approved. 
: a. The LAR Transition Report Section 4.6.2 has been revised to align wi th the approved FAQ 10-0059 and its related closure memo. The revise d Section 4.6.2 provides a description of the process that will be used to assign availability, reliability, and performance goals to High Safety Significant (HSS) SSCs within the scope of the monitoring program. Low Safety Significant (LSS) SSC's do not specifically require assignment of availability, reliability, and performance goals. Programmatic elements such as fire brigade performance, fire watches, combustible controls, etc., will be evaluated using the existing program health process. It is not practical to assign target values of reliability and availability to these attributes so their effectiveness is based on objective and anecdotal evidence evaluated by plant personnel in charge of the fire protection programs as is cu rrently practiced. The revised Section 4.6.2 is provided in Attachment 1 to this enclosure.
: b. The LAR Transition Report Section 4.6.2 has been revised to align wi th the approved FAQ 10-0059 and its related closure memo. The revise d Section 4.6.2 provides a description of how the monitoring program will address response to programmatic elements that fail to meet performance goals. The revised Section 4.6.2 is provided in Attachment 1 of this enclosure.
: c. The LAR Transition Report Section 4.6.2 has been revised to align wi th the approved FAQ 10-0059 and its related closure memo. The revise d Section 4.6.2 provides a description of how    to ULNRC-05876
 
Page 51 of 136 the monitoring program addresses fire protection systems and features and programmatic elements. The revised Section 4.6.2 is provided in Attachment 1 of this enclosure.
: d. As identified in the LAR Transition Report, Attachment A, Table B-1, Section 3.2.3.1, the frequency at which inspections, testing and maintenance of the fire protection systems and features are performed will be evaluated using  EPRI Technical Report 1006756, "Fire Protection Equipment Surveillance Optimization and Maintenance Guide". EPRI Technical Report 1006756 Section 11 contains the following guidance which ensures that reliability levels established are consistent with the FPRA and Maintenance Rule Program: 
"In establishing reliability goals, each plant should determine if other programs, evaluations, or analyses have credited specific reliability values. For example, if the Fire PRA credits a specific level of reliability for a certain suppression system, the target reliability for surveillance optimization should not be below the credited value."
: e. The LAR Transition Report Section 4.6.2 has been revised to align wi th the approved FAQ 10-0059 and its related closure memo. The revise d Section 4.6.2 provides a description of how periodic assessments of the monitoring program will be performed including consideration of internal and external opera ting experience. The revise d Section 4.6.2 is provided in Attachment 1 of this enclosure.
: f. The LAR Transition Report Section 4.6.2 has been revised to align wi th the approved FAQ 10-0059 and its related closure memo. The revise d Section 4.6.2 provides a description of how periodic assessments of the monitoring program will be performed including consideration of internal and external opera ting experience. The revise d Section 4.6.2 is provided in Attachment 1 of this enclosure.
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Page 52 of 136 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 curr ent version cited in Regulatory Guide 1.205, "Risk Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1.
Response to Safe Shutdown Analysis RAI-1 Response provided by ULNRC-05851 dated April 17, 2012.
Safe Shutdown Analysis RAI 02 
 
The nuclear safety capability assessment (NSCA) assumed the loss of instrument air. Please explain how this was incorporated into the initial position of components for circuit analysis. Also, please explain how instrument air failure was consider ed in the non-power operations (NPO) analysis.
 
Response to Safe Shutdown Analysis RAI-2 Response provided by ULNRC-05851 dated April 17, 2012.
Safe Shutdown Analysis RAI 03 Section 4.1.2.2 and Attachment T, Clarification Request 1 of the LAR -NUREG-0830, "Safety Evaluation Report Related to the Op eration 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 requireme nts in the NFPA 805 analysis.
Response to Safe Shutdown Analysis RAI-3 Response provided by ULNRC-05851 dated April 17, 2012.
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Page 53 of 136 Safe Shutdown Analysis RAI 04 Section 4.1.2.2 and Attachment T, Cl arification 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 pr evious approval.
Response to Safe Shutdown Analysis RAI-4 Response provided by ULNRC-05851 dated April 17, 2012.
Safe Shutdown Analysis RAI 05 
 
Section 4.1.2.2 and Attachment T, Cl arification Request 4 of the LA R -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 re duced capacity of le ss than 100 percent.
Response to Safe Shutdown Analysis RAI-5 Response provided by ULNRC-05851 dated April 17, 2012.
to ULNRC-05876
 
Page 54 of 136 Safe Shutdown Analysis RAI 06 Section 4.2.1.2 and Table B-2 of the LAR -To extend the minimum 10 hours coping time, operators must take action to recharge the nitrogen accumulators to support emergency operation of the atmospheric steam dump (ASD) valves and the turbine drive auxiliary feedwater pump (TDAFW) to steam generator (SG) flow control valves. Please explain if the components and/or cables associated with this action are included in the NSCA safe shutdown (SSD) equipment list. Please explain if the steps for recharging the nitrogen accumulators detailed in plant procedures are demonstrated to be feasible.
 
Since the actions to recharge the nitrogen accumulators are not considered recovery actions, please provide a qualitative risk analysis that demonstrates that the risk of failing to perform the actions within the required time frame is low. Should the accumulators not be recharged, please explain if the TDAFW flow control valves can be locally throttled. If so, please explain how these steps are proceduralized and demonstrated to be feasible.
Response to Safe Shutdown Analysis RAI-6 Response provided by ULNRC-05851 dated April 17, 2012.
to ULNRC-05876
 
Page 55 of 136 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 a nd has undergone a focused scope peer review after the fire peer review was comple ted. Please provide the following:
: a. A description of any changes made to the internal events PRA model which are not part of the FPRA and disposition any potential impact on the FPRA results.
: b. A description of the focused scope peer review and disposition any F&Os resulting from this review for their applicability to the current FPRA model.
: c. A discussion of the overall impact of the changes to the internal events PRA model in terms of how the internal events risk profile has changed, that the changes would not impact the FPRA results, and that the internal events PRA model used in the FPRA development can be considered to represent the as-built and operated plant even though additional changes have subsequently been made to the internal events model.
Additional Justification Needed Even if unaffected by fire, new HFEs/HEPs in the internal events PRA model can still affect the fire PRA results because there may be scen arios initiated by fire where non-fire-affected HFEs/HEPs are part of the mitigation. Therefore, item (iii) is applicable to the fire PRA, but it is possible that all such HFEs/HEP s are re-evaluated in light of potential fire effects such that it was determined that there were no changes. If the latter is correct, revise the statement. 
 
The following are questions on the new F&Os in Table 1 (note that this includes Suggestions, to cover the possibility that a Suggestion relative to the inte rnal events PRA could have a greater impact on the fire PRA):
 
(1) F&O 1-7. Can rupture of the RHR or SI system be induced by a fire, including a conditional rupture resulting from a fire-induced initiator?  If so, how?  It appears the F&O cites an underestimate of the rupture probability and, if this is somehow incorporated into the fire PRA, what is the effect on CDF, delta-CDF, LERF and delta-LERF, at least based on a bounding estimate, to ensure the transition conclusions are not affected?
(2) F&O 1-13. If the updated CCF probabilities indicate increases, and these CCFs are
 
part of the fire PRA, what is the eff ect on CDF, delta-CDF, LERF and delta-LERF, at least based on a bounding estimate, to ensu re the transition conclusions are not affected?
 
to ULNRC-05876
 
Page 56 of 136 (3) F&O 4-5. Is FW availability after core damage credited in the LERF model for the fire PRA?  If so, would not this then be applicable?
The responses need to consider not just the FPRA results for the proposed changes which are part of the LAR, but also consider the requested self-approval after implementation of the NFPA 805 license amendment. If appropriate, in order to justify the existing model, pl ease provide sensitivity studies using the updated internal events conditional core damage probabilities.
 
Response to PRA RAI-01
: a. Response provided by ULNRC-05851 dated April 17, 2012.
: b. Response provided by ULNRC-05851 dated April 17, 2012.
: c. The internal events risk profile changed w ith the recent PRA update relative to PRA model revision 4, primarily in that CDF has decreased overall. The three principle reasons are as follows:  i. Addition of a non-safety related motor driven Auxiliary Feedwater Pump ii. The installation of additional offsite power capability from an electrical cooperative substation and the addition of 4 diesel generators from an offsite location iii. Changes to the HRA
 
Items (i) and (ii) were incorporated into the fire PRA (FPRA) prior to submitting the NFPA-805 LAR application. Item (iii) has also been accounted for in the FPRA, because FPRA is required to develop fire specific HEP's, even for the internal events PRA Human Failure Events. As such, the HEPs used in the FPRA ar e specifically re-evaluated in light of potential fire effects and no updates to the FPRA HEPs were deemed necessary in light of the recent internal events PRA HRA updates.
 
As noted, the significant changes to the internal events PRA that have caused a decrease in internal events plant risk have already been reflected in th e FPRA. As such, no significant changes in the FPRA risk results or insight s (i.e. that would a ffect the NFPA-805 LAR application) would be expected if the full extent of the internal events PRA changes were incorporated into the FPRA at this time.
 
Table 1 that follows is a summary of the PWROG focused-scope internal events peer review F&Os and their disposition. All F&Os have been addressed, except where noted.
to ULNRC-05876 Page 57 of 136 PRA RAI 1 Table 1 - PWROG Focused-Scope Internal Events Peer Review F&Os and Their Disposition F&O No. Associated SR(s) (F)inding or (S)uggestion Brief Description Disposition for IE PRA Disposition for FPRA 1-1 IC-C5 F Convert SSIE and ISLOCA IE frequencies to Rx-yr basis.
 
A plant availability factor of 0.9, used for the other internal events initiators, was applied to the SSIE and ISLOCA frequency
 
quantification. FPRA calculated its own IEFs. [n/a] 1-2 IE-C8 F Assess alternate alignments for loss of SW
 
and CCW.
Calculations EA-05 Revision 1 and EG-19 Revision 1 were modified to provide more in-
 
depth discussion and justification for the use of a
 
single alignment. FPRA models multiple alignments for ESW and CCW and uses split fractions for percent of time spent in each one.
[n/a]  1-4 IE-C10 F More than one initiator BE exists in SSIE and ISLOCA cutsets.
The ISLOCA and Loss of All Service Water models were
 
revised such that cutsets now contain only one initiator/frequency BE. Does not apply. FPRA has its own initiators and calculates its own ISLOCA frequency. [n/a]
1-7 1 IE-C14 F Rupture probability of RHR and SI systems
 
needs to be based on failure probability of all piping/components (not on the weakest location).
The RHR and SI system rupture probabilities used in ZZ-138, Rev. 0, Add. 1 were revised in response to this F&O. For both systems, a summation of the piping/component rupture probabilities is now used.
Therefore, this F&O has been addressed. FPRA uses overpressure probabilities from ZZ-138. The updated probabilities in ZZ-138, Rev 0, Addenda 1 have not been incorporated into the FPRA as of March 2012. These will be updated with next FPRA update. Fire-induced rupture [due directly to pipe damage] of RHR or SI piping is not considered in the Fire PRA. 1-8 IE-D1, AS-C1,  LE-G1, IFSN-A12, IFSN-B1 F Documentation builds on earlier documentation.
 
The three examples cited in the F&O were addressed.
Not applicable to the FPRA. to ULNRC-05876
 
Page 58 of 136 F&O No. Associated SR(s) (F)inding or (S)uggestion Brief Description Disposition for IE PRA Disposition for FPRA 1-9 IE-A5 F Is loss of ESW a separate IE?  Further evaluation indicates that creation of a separate Loss of ESW initiating event is not necessary or justified. Finding is not applicable to the FPRA. 
 
Loss of ESW/SW is considered in fault trees for FPRA.
1-14 1 DA-D3,  DA-E2 F Treatment of CCF uncertainty.
F&O was evaluated. F&O response includes recommendation to perform CCF uncertainty sensitivity runs in the future once the Data Parameter file is completed. Internal Events PRA response is
 
applicable to FPRA. 1-20 IFSO-A4, IFSO-B2, IFEV-A7 F Either apply applicable generic data for human-induced flooding or develop plant-specific human-induced flood frequencies. Resolution of this F&O is pending. This F&O is related to the internal events IF analysis, and does not impact the Fire PRA. The flooding events referred to in the F&O are not used in the FPRA. [n/a]
1-25 1 AS-B3 F Consideration of phenomenological
 
conditions
 
Resolution of this F&O is pending. It is not anticipated that resolution of this F&O would have any impact on the IE
 
PRA. Resolution of this F&O is pending. It is not anticipated that resolution of this F&O would have any impact on the
 
FPRA. 2-6 LE-B1 F Probability of successful ex-vessel cooling should be justified.
MAAP runs were performed to justify the probability used.
The FPRA Level 2 evaluation uses probabilities and split fractions from the previous Level 2 analysis. [n/a]    to ULNRC-05876
 
Page 59 of 136 F&O No. Associated SR(s) (F)inding or (S)uggestion Brief Description Disposition for IE PRA Disposition for FPRA 3-1 AS-A1, AS-A2,  AS-A3 F The F&O questions whether the S(3) event tree should question loss of RCP seal cooling.
The need to question seal cooling following a very small LOCA was evaluated. It was determined that loss of seal cooling does not need to be included in the S(3) event tree. S3 tree for FPRA is not the same event tree as for the internal events. S3 event tree for FPRA delineates very small LOCA's caused by fire induced events.
Loss of seal cooling caused by 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 relate events, but does not postulate a random S3 LOCA simultaneous with a fire induced S3 LOCA. [n/a] 3-6 AS-A5, AS-B2 F Include potential for consequential LOOP in RCP seal LOCA AS analysis.
This F&O was evaluated. As a result, consequential LOOP was added to the Tc and Tsw event trees. Residual LOSP and consequential LOSP are not included in FPRA. (A LOSP resulting from a fire 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]
1-3 1 IE-C8 S Include PEG01A FTS on loss/recovery of power to the pump.
 
Calculation EG-19 Revision 1 was modified such that pump start failures (including common cause start failures) for PEG01A (the running pump) are included for loss of normal power to the running pump followed by power recovery. CCF failures on LOSP are included in the FPRA. A secondary fail-to-start failure mode is not included for the 50% of the time pump A is running. 1-5 IE-C9 S This F&O questions the mission time used for CCW pressure transmitters in the CCW initiator model/FT. This F&O has not been resolved. This issue pertains only to the Loss of All CCW initiator fault tree/quantification. It does not impact
 
the FPRA. to ULNRC-05876
 
Page 60 of 136 F&O No. Associated SR(s) (F)inding or (S)uggestion Brief Description Disposition for IE PRA Disposition for FPRA 1-6 IE-C14 S Justify the 1" ISLOCA screening criterion.
ZZ-105, Rev. 0, Add. 1 was
 
revised to add justification for the 1" screening criterion (i.e.,
"e") used in the ISLOCA
 
location review. Same justification as for Internal Events
 
PRA. 1-12 DA-D6 S Suggested data documentation enhancements.
 
Additional information was added to the affected documentation. No fire response necessary. [n/a]
1-13 1 DA-D6 S Two potential issues identified with application of the common-cause
 
data. The identified issues were addressed.
The CCF values and basic events for
 
the Internal Events PRA were updated in 2011. The work was independently peer reviewed in August 2011 and peer review comments were addressed and finalized in October 2011, well after the NFPA-805 LAR was submitted. Some CCF values increased and some
 
decreased. A succinct sensitivity study for each CCF value to determine its individual effect has not been performed at this time. The effort to
 
incorporate each of the updated CCF values in the FPRA is substantial, as it would require re-quantification of all fire scenarios and all Fire Risk Evaluations. The updated CCF's are scheduled to be incorporated as part of the next FPRA revision.
1-15 1 LE-C6,  LE-C7 S The F&O suggests consideration of pre-
 
initiator CTMT isolation
 
failures in the CTMT isolation systems model.
This F&O has not yet been addressed. However, the LERF analysis already includes a "FAIL_LEAK" event, obviating the need to take any action in response to this F&O. This F&O is a suggestion, which is not yet addressed by the Internal Events PRA. Consequently, it has not been addressed by the FPRA. to ULNRC-05876
 
Page 61 of 136 F&O No. Associated SR(s) (F)inding or (S)uggestion Brief Description Disposition for IE PRA Disposition for FPRA 1-16 IFSN-A6 S Need qualitative assessment of pipe whip, humidity, temperature, etc., in IF analysis.
To address this F&O, documentation was added to the Internal Flooding Notebook. Random flooding events are not
 
addressed in the FPRA. 1-18 AS-A4 S Suggestion for AS documentation enhancement. F&O response clarifies Callaway's current approach to documentation. Not applicable to FPRA documentation report. 1-19 IFSO-B2, IFSN-A15 S Flood source screening documentation enhancement. Information added to the IF Notebook. Random flooding events are not
 
addressed in the FPRA.
1-23 IFQU-A3, IFQU-B3 S Suggested IFQU documentation enhancement relative to
 
screening quantification
 
decisions. Minor revisions were made to
 
the IF Notebook to address this F&O. Random flooding events are not
 
addressed in the FPRA. 1-26 IFEV-A6 S Provide a more complete discussion of plant-specific experience that could impact flood
 
likelihood. This F&O has not yet been
 
addressed in the IE PRA. Random flooding events are not
 
addressed in the FPRA. The F&O is related to the Internal Events PRA IF analysis, and has no bearing on the
 
FPRA. 2-1 LE-A3 S Suggested LERF documentation enhancement. Additional information added to LERF Notebook.
Internal Events LERF documentation is not applicable to FPRA. 2-2 DA-A2 S Suggested addition of component boundary, failure mode and success criteria discussion to DA documentation. Additional discussion added to Data calculation. This is incorporated into the FPRA by virtue of the fact that the random failure probabilities in the FPRA are from the Internal Events PRA. 2-3 IFPP-B1 S Documentation suggestion relative to IFPP. Minor revisions made to the IF Notebook to address this F&O. Random flooding events are not
 
addressed in the FPRA. 2-4 LE-C1 S More justification required for the definition used for early release. Additional information added to the LERF Notebook.
Not applicable to the FPRA. to ULNRC-05876
 
Page 62 of 136 F&O No. Associated SR(s) (F)inding or (S)uggestion Brief Description Disposition for IE PRA Disposition for FPRA 2-5 LE-C5 S Suggestion relative to the use of conservative versus realistic LERF success criteria. Specific suggestions of the F&O
 
were addressed. F&O not applicable to FPRA LERF analysis. 2-8 IFSO-A5 S Suggestion to add temperature of flood sources to IF documentation. Information added to IF Notebook in response to this F&O. Random flooding events are not addressed in the FPRA. 3-2 AS-A2 S Revise ZZ-275 CSF/SC tables to match current Tc and Tsw event trees. Revised tables were added to the AS calculation set to address this F&O. Internal Events PRA event tree documentation not applicable to FPRA. 3-5 AS-A7 S Suggestion to consider the need to add the potential for a stuck-open PORV to T1s, Tc and Tsw event trees. Suggestion was evaluated, and justification for not adding a stuck open PORV to these ETs was generated. Spurious-open PORV is a consequential event in the FPRA, which will appear in any scenario where fire damage can cause it to occur. [n/a]
3-8 IFQU-A1, IFQU-B2 S Suggestion for additional documentation. This suggestion has not yet been addressed. However, it pertains to the IFQU element, and does not impact the Fire PRA. Random flooding events are not addressed in the FPRA. 3-9 IFSO-A1, IFSN-A8 S Document a basis for floor penetrations and block walls not failing due to flood loads.
 
This suggestion has not yet been addressed. However, it pertains to the IF analysis, and does not impact the Fire PRA. Random flooding events are not
 
addressed in the FPRA. 3-10 IFSN-A10 S Suggestion to consider the potential for floor drain
 
blockage.
 
This suggestion has not yet been addressed. However, it pertains to the IF analysis, and does not impact the Fire PRA. Random flooding events are not addressed in the FPRA. 4-1 LE-C1 S Suggested minor revision to LERF-related text. Minor revision made to LERF Notebook to address this F&O. F&O is not applicable to the FPRA. to ULNRC-05876
 
Page 63 of 136 F&O No. Associated SR(s) (F)inding or (S)uggestion Brief Description Disposition for IE PRA Disposition for FPRA 4-2 LE-C8 S Suggestion for additional discussion of how Level 1 and Level 2 models are linked. Information added to LERF Notebook. F&O is not applicable to the FPRA. 4-3 AS-A7, AS-A10 S Suggested enhancement of RCP seal LOCA accident sequence documentation. Additional text justification was developed, and will be included in an AS Notebook, currently under development. The event tree assumptions that are
 
critiqued here are not made in the FPRA. F&O does not apply to the FPRA. 4-4 LE-G2 S Suggestion relative to LERF documentation enhancement. Information added to LERF Notebook. F&O is not applicable to the FPRA. 4-5 LE-C9 S Suggestion to justify feedwater availability
 
after core damage (as
 
credited in the LERF model). Justification/text added to the LERF Notebook. F&O is not applicable to the FPRA because the status of feedwater availability at the time of core damage or after core damage is not considered in the FPRA LERF model.
Note 1 - These F&Os will be evaluated and/or incorporated into the fire PRA during the next update, per Implementation Item 12-805-001 shown in the updated Attachment S to this enclosure.
to ULNRC-05876 Page 64 of 136 Probabilistic Risk Assessment RAI 02 The peer review description addresses the relevant internal events PRA sta ndard, 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 a nd 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 qualif ications and indicate how any id entified gaps were dispositioned.
This also applies to the FPRA p eer review. In your response, pl ease address both peer reviews.
Response to PRA RAI-02 Response provided by ULNRC-05851 dated April 17, 2012.
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Page 65 of 136 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 dis position any peer review F&Os.
Additional Justification Needed  The response seems to say that a copy of the LERF event tree was used only to simplify linking between the trees. However, in the response to an audit question, it is explained that a simplified version of the LERF event tree was used, what was simplified, and why this is OK. This more descriptive response should be includ ed to the actual RAI 3 response.
 
Response to Probabilistic Risk Assessment RAI-03 The Callaway Plant FPRA developed a LERF model specifically for the FPRA from the LERF model for the Internal Events PRA. The development was necessary to enable the FPRA LERF model to run from the same set of event trees as the Level 1 fire sequence development and run from the same PRA batch file as the Level 1 FPRA core damage sequences. 
 
This model was complete and in place at the tim e of the FPRA peer review in October 2009. Although there are no specific Fire Supporting Requirements for LERF, deficiencies in the LERF model would be identified from the following Fire SR's, which relate back to Internal Events SR's from Chapter 2 of the combined PRA Standard: 
 
PRM-B1 PRM-B14 PRM-B15 PRM-C1 FQ-D1 FQ-E1 FQ-F1 UNC-A2  A review of the F&O's from the Callaway Plant FPRA peer review shows 2 F&O's which cited LERF Findings. These are shown in the table below, along with their resolution.
to ULNRC-05876
 
Page 66 of 136 SR DESCRIPTION OF F&O RELATED TO LERF SUGGESTION OR FINDING RESOLUTION PRM -C1  Documentation needs to be expanded. The following are examples of documentation
 
that should be provide d to support the Plant Response Model (PRM) and quantification: 
 
(8) Document the linkage of the containment
 
isolation in the LERF calculation. Finding Documentation was expanded in section
 
4.8.1 of the fire-induced risk model
 
report (17671-004) to
 
address this F&O. FQ -F1  Documentation was limited for fire quantification. 
 
(5) Document the significant contributors to the Fire PRA. Document both the dominant
 
LERF and CDF contributors. Finding Documentation was expanded in sections
 
5.2 and 5.3 of the integrated fire risk
 
report (17671-013) to
 
address this F&O.
The FPRA developed a LERF model ba sed on the 2006 version of the Ca llaway Plant Internal Events PRA. The 2006 Internal Events PRA LERF analysis develops a LERF equation which considers a) core-containment energetics, b) ISLOCA, c) steam generator tube rupture, d) containment isolation failure and leakage. The Internal Events PRA uses a containment event tree to develop a LERF split fraction representing LERF due to containment failure from core - containment energetics. Items (b), (c), and (d) from above are modeled explicitly. 
 
The FPRA used the same process for LERF model development, except that LERF split fraction due to core-containment energetics for part (a) above ar e derived directly from the Internal Events PRA, based on similarity of Plant Damage States (PDS). Potential PDS's for the fire sequence were compared to the PDS's from the Internal Events PRA sequences and the maximum LERF split fraction for any PDS resulting from any applicable Internal Events PRA PDS was used in the FPRA. Items (b), (c), and (d) from above were explicitly calculated for the FPRA given the applicable specific failures in the fire scenario. 
 
The Internal Events PRA was recently updated. The LERF model for the Internal Events PRA was updated and finalized in December 2011, after the NFPA 805-license amendment request was submitted to the NRC. The FPRA LERF model will be updated to be consistent with the Internal Events PRA LERF model phenomenology, split frac tions, and probabilities during the next FPRA update per Implementation Item 12-805-003. Implementation Item 12-805-003 is shown in the updated Attachment S to this enclosure.
 
Overall, the LERF / CDF split fraction in the current Internal Events PRA model is 1.3% (i.e., 2.29E-5 CDF / 3.09E-7 LERF) versus a LERF ratio of 1.97% for the FPRA (2.03E-5 CDF / 3.99E-7 LERF). In accordance with RG 1.205 and RG 1.174, LERF does not become the determining metric for an application unless the LERF / CDF split fraction it is greater than 10%. The LERF fractions for the Callaway Plant FPRA and the Internal Events PRA are similar and are both significantly less than
 
10%.
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Page 67 of 136 Probabilistic Risk Assessment RAI 04 The resolution of a number of F&Os from the intern al events review does not appear to fully address the impact of the resolution on the NFPA-805 resu lts. Please justify the proposed resolution as follows:
: a. During the audit, the licensee identified that F&O SY-2 dis position 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 cond ucted. 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 revise d 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 revi ew. 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 ga ps. 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 us e 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 inte grity of the piping (i.e., HRR levels assumed for transient combustible ignition sources in the analyses performed to address the use of this polyethylene piping.)
Additional Justification Needed:
"No Storage" and "No Hot Work" are cited as the basis for assuming a transient combustible HRR of 69 kW (98th %ile) in Area C-1, containing high-density polyethylene piping. In response to PRA RAI 23, multiple reasons were cited as the basis for the lower    to ULNRC-05876
 
Page 68 of 136 HRR assumption (69 kW). Which of these other factors are also applicable in Area C-1, (i.e., not just the designations of "No Storag e" and "No Hot Work"). Also, the response cites changes made to the Transition Report (LAR) reflecting an updated analysis for Area C-1, specifically Table 4-3, Att. C, Att. D and Table W-2.  (Submit LAR update)
Response to Probabilistic Risk Assessment RAI-04
: a. Response provided by ULNRC-05851 dated April 17, 2012.
: b. Response provided by ULNRC-05851 dated April 17, 2012.
: c. Response provided by ULNRC-05851 dated April 17, 2012.
: d. Response provided by ULNRC-05851 dated April 17, 2012.
: e. Response provided by ULNRC-05851 dated April 17, 2012.
: f. Background 
 
Callaway Plant reported as a Li censee Event Report (LER), a c ondition identified in the plant involving High Density Polyethylene (HDPE) piping that affects the NFPA 805 License Amendment Request (LAR) and its supporting documentation (reference LER 2011-006-00 transmitted via ULNRC-05836 dated 1/6/12). During a review of the analysis associated with
 
Fire Area C-1, Pipe Space and Tank Area, Control Building Elev. 1974' / 1984' it was determined that the HDPE piping that had been installed by plant modification could be affected by a fire. The resulting HDPE pipe fail ure could create a floodi ng condition where one train of required Essential Service Water (ESW) equipment is not maintained free of fire damage. Fire Area C-1 (CB 1974' room 3101) contai ns both trains of ESW supply and return piping. Fire Protection Program compliance in Fire Area C-1 is met by the ESW trains being separated by 20 feet with no intervening combustibles and automatic detection and suppression.
 
NFPA 805 LAR Impacts HDPE Piping During development of the NFPA 805 LAR and the associated Nuclear Safety Capability Assessment (NSCA) for fire area C-1, the impact of a fire on the HDPE pipe was not considered. Fire Area C-1 is comprised of two fire zones (rooms), room 3104 which is a stairwell and room 3101 which is the large pipe space that contains the ESW system related equipment. Fire Area C-1 was determined to be deterministically compliant based on the two trains of ESW piping and valves to be adequately separated by 20 feet of separation with the presence of automatic detection and suppression in Room 3101. Additionally, fire modeling was performed in Room 3101 and as a result of FPRA risk insights, fire area C-1 Room 3101 had been designated a "No Storage Location" and "No Hot Work Location". The only ignition sources in Room 3101 are transient hot work and transient combustibles. To address the fire    to ULNRC-05876
 
Page 69 of 136 induced HDPE pipe failure scenario, two new variances from the deterministic requirements (VFDR's) were developed for Fire Area C-1; one for each ESW train of HDPE piping in Room 3101. A Fire Risk Evaluation (FRE) was completed for Fire Area C-1 to evaluate the impact of the new VFDR's and the results are documented in a revised C-1 Fire Safety Analysis (FSA), calculation KC- 113, Fire Safety Analysis for Fire Area C-1". To support the FRE, fire modeling was conducted in fire area C-1 using Fire Dynamics Simulator (FDS) to evaluate the impact of transient fire scenarios in the immediate vicinity of the HDPE pipe for their ability to damage the pipe. Because Fire Area C-1 is considered a "No Storage" and "No Hot Work" area, the HRR postulated for the transient fires is 69 kW. C-1 is a pipe chase with limited equipment and large combustible liquid fires are not expected. Since only small quantities of trash in temporary containers can be expected, a 69kW peak heat release rate was determined to be appropriate to represent this quantity of combustibles. The 69kW heat release rate bounds the small trash can fires reported in NUREG/CR-6850 Appendix G. The FDS fire modeling demonstrated that the HDPE pipe will remain free of fire damage. The FRE also evaluated defense in depth and safety margin and determined a Main Control Room operator action was required to align ESW valves which ensures that they are in the required NSCA position should flooding occur.
A new delta risk calculation was developed for Fire Area C-1, calculation 17671-FRE-C-1, "Fire Risk Evaluation for Fire Area C-1". Because the VFDR's (pipe) are not damaged by any fire scenario damage set, the delta risk for the VFDR's is negligible and overall absolute fire risk in Room 3101 remains unchanged.
 
The Callaway Plant Transition Report has been revised to change the NFPA 805 regulatory basis for fire area C-1 from 4.2.3.2 to 4.2.4.2 in Transition Report Table 4-3. In addition, Transition Report Attachments C, D and W, have been revised. The VFDR's have been added to the Fire Area C-1 discussion in Attachment C, "NEI 04-02 Table B-3 Fire Area Transition", Attachment D "Non Power Operational Modes Transition" and Attachment W, "Fire PRA Insights". These changes are reflected in Attachments 1, C, D and W of this enclosure. Additionally, background documents have been updated to reflect the VFDR's and other required description changes as follows;
* Calculation KC-113, "Fire Safety Analysis fo r Fire Area C-1" - Re vised to include a description of HDPE pipe, to add the VFDR's, and revise the fire area boundary description to include the prefabricated enclosure used at the ESW pipe wall penetration.
* Calculation KC-116, "Fire Safety Analysis fo r Fire Area C-5" - Re vised to reflect the change to the description of the fire barrier interface with C-1.
* Calculation KC-117, "Fire Safety Analysis fo r Fire Area C-6" - Re vised to reflect the change to the description of the fire barrier interface with C-1.
Calculation KC-57, "Detailed Fire Modeling Report for Fire Area C-1" - Revised to include a description of HDPE pi pe, to revise the fire area boundary description to include    to ULNRC-05876
 
Page 70 of 136 the prefabricated enclosure used at the ESW pipe wall penetration and to include a discussion of the FDS fire modeling performed.
* Calculation KC-58, "Detailed Fire Modeling Report for Fire Ar ea C-5" - Revised to reflect the change to the description of the fire barrier interface with C-1.
* Calculation KC-59, "Detailed Fire Modeling Report for Fire Ar ea C-6" - Revised to reflect the change to the description of the fire barrier interface with C-1.
* Calculation R1984-001-001, "Fire Dynamics Simulator (FDS) Analysis To Support Detailed Fire Modeling" - Revised to add the FDS fire modeling which evaluated transient fire impacts to the HDPE pipe.
* Calculation KC-26, "Nuclear Safety Capability Assessment" - Revised to address the VFDR's and their resolution.
* Fire Pre-Plan Manual- Revised fire res ponse guidance for C-1 to identify HDPE piping locations.
* Calculation 17671-FRE-C-1, "Fire Risk Eval uation for Fire Area C-1"- New document Elastomeric Isolation Joints During an extent of condition evaluation for the HDPE pipe issue described above, additional elastomeric components were identified in the plant that if failed by fire could have a potential for adverse consequences on protected train equipment. In Fire Areas D-1 and D-2, the two emergency diesel generator (EDG) rooms, there is ESW piping which contains elastomeric expansion joints. Should an elastomeric joint fail due to fire e xposure the resultant flooding in the fire area could impact adjacent fire areas and affect redundant train equipment if no operator response is taken. Each EDG is cooled by ESW, which has both supply and return piping containing elastomeric expansion joints, that if failed due to fire damage will result in ESW flooding the EDG room.
For Fire Area D-1 if no action is initiated to isolate the ESW water fl ow, the resultant flooding could impact the adjacent fire area which contains the credited train of electrical switchgear. During development of the NFPA 805 LAR and the associated NSCA analysis, Fire Areas D-1 and D-2 were determined to be deterministically compliant with the requirements of NFPA 805 Chapter 4. The elastomeric components failures were evaluated and determined not to meet the criteria of a VFDR as the failed condition is recovered by a Main Control Room action. The Callaway Plant LAR is not affected; however, LAR background documents have been revised as follows;  Calculation KC-149, "Fire Safety Analysis fo r Fire Area D-1" - Revised to include a description of the expansion joints and the impact.
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Page 71 of 136 Calculation KC-150, "Fire Safety Analysis fo r Fire Area D-2" - Revised to include a description of the expansion joints and the impact.
Calculation KC-75, "Detailed Fire Modeling Report for Fire Ar ea D-1" - Revised to reflect the change to the damage sets for fire scenarios that are in the vicinity of the elastomeric joints.
Calculation KC-76, "Detailed Fire Modeling Report for Fire Ar ea D-2" - Revised to reflect the change to the damage sets for fire scenarios that are in the vicinity of the elastomeric joints.
Calculation KC-26, "Nuclear Safety Capability Assessment" - Revised to address a main control room action to isolate ESW.
 
Fire Pre-Plan Manual- Revised fire response guidance for fire area D-1 and D-2 to identify that elastomeric joints ma y create flooding conditions.
OTO-KC-00001, "Fire Response", Addendums D-1 and D Revised to add a Main Control Room step to secure the affected train ESW pumps to mitigate the flooding.
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Page 72 of 136 Probabilistic Risk Assessment RAI 05 If changes to the FPRA model have been made subsequent to the completion of the peer review of the FPRA, please provide a description of any new models or methods that have been implemented for the FPRA, including any subsequent focused scope peer reviews of the models or methods.
Response to Probabilistic Risk Assessment RAI-05 Response provided by ULNRC-05851 dated April 17, 2012.
 
Probabilistic Risk Assessment RAI 06 The resolution of a number of F&Os from the FP RA review does not appear to fully address the impact of the resolution on the NFPA-805 results. Pl ease 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 pre ssurized water reactor (PWR) MSO list" was reviewed. Pl ease 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 clarif ying 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 spur ious 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 cabin et, but the response does not specifically address a fire originating in the cabinet. The second part, the potential comple xity of a fire event causing    to ULNRC-05876
 
Page 73 of 136 spurious safety injection (SI) and containment isolation, is not addressed in the disposition.
Please provide a more complete disposition of this F&O.
Response to Probabilistic Risk Assessment RAI-06 Response provided by ULNRC-05851 dated April 17, 2012.
 
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Page 74 of 136 Probabilistic Risk Assessment RAI 07  A number of issues with the Human Reliability Analysis (HRA) need to be clarified: 
: a. F&O HRA-E1 The F&O indicates that the human error analysis credits instrumentation not traced to assure availability. Please confirm that credited instrumentation relied upon for the HRA is based on availability of instruments free from fire damage. Additional Justification Needed:
While it likely may be inferred, the RAI requested confirmation that available instruments were free from fire damage. The response cites the list of available instruments but does not explicitly state that they are free from fire damage when being credited. Provide this confirmation if correct. If not, explain the basis for taking credit.
: b. Please provide the basis for assuming a screening human error probability of 0.1 for failure of successful operation at the Auxiliary S hutdown 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. Pleas e include the results of a sensitivity evaluation if each human error probability is assumed to be 1.0 (or some other bounding value, with justifica tion), or provide the basis for the assumed value being appropriate.
Additional Justification Needed Do the calculational results (Tables 2 through 4) reflect removal of credit for CPTs in all scenarios?  If not, how would the results change if this credit were removed entirely?  Note that, for the delta calculations, the credit should be removed in both the base and comparative cases.
: d. F&O FQ-C1 The F&O identifies that the HRA dependency analysis does not consider
 
execution dependencies for local actions for fire scenarios. This item is indicated to be closed, but the disposition is to review a nd disposition these dependencies in the next    to ULNRC-05876
 
Page 75 of 136 FPRA update. Please confirm completion of this item sufficient to resolve the technical issue for the existing fire PRA used to support this application.
: e. Conservatism in the current state of FPRA was cited as the basis for: (1) considering it premature to perform a detailed dependency analysis for the fire HRA; (2) dismissing completeness uncertainty as a current concern in fire HRA; (3) not performing uncertainty analysis on fire risk and delta-risk. Please provide either: (1) sensitivity evaluations to address the potential impact of not explicitly addressing these issues or (2) a discussion of the plant-specific aspects of the FPRA for Callaway that constitute the basis for the cited
 
conservatism.
Additional Justification Needed Table 3-1 of Calculation 17671-014 is cite d as providing discussion as to why completeness uncertainty does not apply to fire HRA. State explicitly this material from Table 3-1. In addition, Table 4-3, presumably of the same calc ulation, is cited as providing importances of applicable recovery actions. However, there appears to be no such table in the calculation. If Table 4-3 of the LAR is m eant, note that this does not address HRA, but rather Fire Protection Systems and Features. Provide clarifica tion and, if necessary, correction.
: f. If a sensitivity/uncertainty analysis was performed for the Fire LERF and Delta-LERF (LERF) after the LERF model was ready, please report the results. If not, please perform an analysis or justify the basis for assuring that the insights to be gained from a sensitivity/uncertainty analysis were obtained otherwise and the means of doing so.
Additional Justification Needed "Report the results" means that they should actually be docketed, either in this RAI response or as part of an update to the LAR, not just referenc ed as available in a portal document (17671-014, App. B). This includes th e results for CDF, LERF, delta-CDF and delta-LERF. Provide, e.g., one of the following: (1) add the materi al from the portal document to the RAI response, or (2) embed this material in the LAR as updated. Keep in mind the potential effect of the response to RAI 9b on the material in 17671-014, App. B (Sensitivity #1).
Response to Probabilistic Risk Assessment RAI-07 a) F&O HRA-E1-1 (Suggestion level F&O) was wr itten in October 2009 dur ing the Fire Peer Review. Since 2009, the process described below was implemented to ensure the human reliability analysis only credited instruments that are free of fire damage. 
 
To start, Table 3-3 of the Callaway Plant Fire HRA repor t (Calculation 17671-011) provides the specific instruments that are required by operator actions in order to accomplish diagnosis. Each of the instruments listed in Table 3-3 was cable traced. Table 3-3 was updated following the 2009 FPRA peer review to include references to specific instruments. to ULNRC-05876
 
Page 76 of 136
 
Next, in order to ensure the availability of instrumentation for operator actions, human failure events (HFE's) were grouped into two categories. 
: 1) HFE's directed by the fire response proce dures and evaluated as part of the NSCA (Calculation KC-26). These HFE's are show n in Table 3-4 (Local Actions) and 3-5 (Control Room Actions) of the Callaway Plant Fire HRA report (Calculation 17671-011). Part of the NSCA evaluation of human actions is to ensure adequate instrumentation and cues are available to support the action on an area-by-area basis wherever the action is required. For these HFE's, the PRA used the NSCA analysis to ensure that instrumentation and procedur es are available to support the action. 
: 2) HFE's directed by the internal events Emergency Operating Procedures (EOP's). Table 3-3 of the Callaway Plant Fire HRA repor t (Calculation 17671-011) lists the HFE's associated with this category and provides the specific instruments that are required for diagnosis. Each of the instruments li sted in Table 3-3 was cable traced.
Instrument Availability Considerations During Quantification:
Two trains of instrumentation are typically provided for any operation. Human error probabilities (HEP) values are developed for three cases of instrumentation availability:
a) Both trains of instrumentation available. HEP at a nominal value that accounts for fire-effects.
b) One train of instruments available. HEP increased in accordance with guidance in NUREG-1921.
c) No instrumentation available. HEP is guaranteed failed. 
 
In order to more efficiently conduct the fi re scenario quantific ation while providing a conservative basis for the HEP's, HEP's were calculated as case (b) crediting only a single train of instrumentation. This is a higher HEP value than the nominal fire value, but precludes the needs to match a specific HEP with every fire scenario. For all instruments in Table 3-3, it was verified that at least one train of instruments were available for each fire area, so that condition (b) applied. If this was not the case, the HEP was assigned a value of 1.0.
b) Response provided by ULNRC-05851 dated April 17, 2012.
 
c) The human reliability analysis (HRA) methodology used to calculate the human error probabilities (HEPs) for the three cited events was consistent with the method that was applied to all human failure events (H FEs) in the Callaway Plant FPRA. This method was examined during the October 2009 Peer Review and there are no HRA-related peer review comments that are open.
 
There are three HFEs involved in this question.
These HFEs are shown in Table 1 below, along with the Fire Areas in which they are credited, and their nominal HEPs.
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Page 77 of 136 Table 1 - Applicable Human Failure Events Event Description Credited Fire Areas Nominal Probability OP-OMA-FF-EGRVAB Operator fails to locally close spuriously open CCW surge tank vent valve for
 
CCW train swap. C21 / C22 / C271.3E-1 OP-OMA-FF-ISOEG Operator fails to locally close spurious open CCW/ESW fill  valves.
A21 / C21 /
C22 / C24 /C27 5.0E-1 OP-OMA-FO-RCPTRP Operator fails to locally trip the RCPs when MCR trip capability is failed.
C27 2.9E-1 A sensitivity study was performed to find the CDF/LERF/delta CDF/delta LERF changes if no credit is allowed for the HEP's in Table 1. This is a bounding case.
 
In addition, in order to combine sensitivity studies, the changes were done for the case with and without credit for the control power transformer (CPT) in determining the probabilities for spurious valve actuation (issue from RAI-PRA-9b). The existing License Amendment Request (LAR) results, which credit these HEPs at their nominal values, and the sensitivity results are compared in Tables 2a and 2b below.
 
Note that all results tables have an "a" and a "b" version. The "a" versions show risk results that credit control power transformers (CPT) during the calculation of spurious actuation probabilities based on fire-induced cable damage. The "b" version of each table show risk results with the CPT credit removed.
 
Table 2a - LAR vs. HEP Sensitivity Results (with CPT credit)
LAR Results HEP = 1.0 Sensitivity Results CDF (per yr.)
LERF (per yr.) CDF (per yr.) LERF (per yr.) CDF (per yr.)
LERF (per yr.) CDF (per yr.) LERF (per yr.)
8.07E-06 2.50E-07 3.87E-06 2.03E-07 8.45E-06 2.86E-07 4.22E-06 2.38E-07 Table 2b - LAR vs. HEP Sensitivity Results (without CPT credit)
LAR Results HEP = 1.0 Sensitivity Results CDF (per yr.)
LERF (per yr.) CDF (per yr.) LERF (per yr.) CDF (per yr.)
LERF (per yr.) CDF (per yr.) LERF (per yr.) 2.29E-05 5.04E-07 6.24E-06 2.77E-07 2.35E-05 5.41E-07 6.78E-06 3.14E-07 to ULNRC-05876
 
Page 78 of 136 The increase in each metric is shown in Tables 3a and 3b.
Table 3a - Risk Metric Increases (with CPT credit)
CDF  (per yr.)
LERF (per yr.) CDF (per yr.) LERF (per yr.)
3.79E-07 3.55E-08 3.49E-07 3.48E-08
 
Table 3b - Risk Metric Increases (without CPT credit)
CDF  (per yr.)
LERF (per yr.) CDF (per yr.) LERF (per yr.) 5.80E-07 3.79E-08 5.41E-07 3.70E-08
 
The increases in CDF and LERF are compared to the plant totals in Tables 4a and 4b.
Table 4a - Delta Risk Increases Relative to Plant Totals and  Regulatory Limits (with CPT) Metric LAR RAI 07c Increase Total Limit CDF (per yr.) 1.87E-06 3.49E-07 2.22E-06 < 1E-5 LERF (per yr.) 3.84E-08 3.48E-08 7.32E-08 < 1E-6 Table 4b - Delta Risk Increases Relative to Plant Totals and  Regulatory Limits (without CPT) Metric LAR RAI 07c Increase Total Limit CDF (per yr.) 1.87E-06 5.41E-07 2.41E-06 < 1E-5 LERF (per yr.) 3.84E-08 3.70E-08 7.54E-08 < 1E-6
 
This calculation has taken the HEPs in questi on and bounded their uncertainty by setting them to 1.0 in the sensitivity quantification. Although the HRA methods us ed to calculate the nominal HEP values were reviewed during the FPRA Peer Review and are consistent with those of the rest of the Fire PRA, this analysis has been performed to show the "worst-case" scenario.
 
As shown in Tables 4a and 4b, even with this bounding approach, if the delta risk impact of this sensitivity study is added to the existing LAR delta risk totals, the resulting totals are still comfortably below the regulatory limits. Additiona lly, the final risk insights shown in Tables 4a and 4b indicate that the conclusions of this sensitivity study are not significantly impacted by removing credit for the CPTs in the spurious actuation probability ca lculations. The results without CPT credit in Table 4b show a slightly higher increase in delta risk compared to the CPT-credited results in Table 4a, but both sets of results maintain acceptable margin to the regulatory limits. 
 
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Page 79 of 136 The nominal HEP values used for the LAR risk calculations were calculated with the standard FPRA HRA methods used throughout the model, wh ich were reviewed du ring the Peer Review in October 2009. However, even if the HEPs ar e set to 1.0 to bound potential uncertainty in the HEP values, the overall risk insights are unaffected.
 
d) Response provided by ULNRC-05851 dated April 17, 2012.
 
e) The wording of the Callaway Plant fire HRA report could have been better structured to convey the detailed dependency analysis conducte d as part of the fire HRA. The Callaway Plant Fire HRA conducted a detailed dependency anal ysis as described in the response to PRA RAI 7d. The peer review observation of the Callaway Plant HRA dependency analysis was a Suggestion level F&O to conduct a review and ensure that comb inations of execution HFEs that occur in the FPRA cutsets are actually independent. A revi ew of the cutsets was conducted (by visual inspection) during the Fire Risk Evaluation process that examined each risk-significant fire area. Since the Callaway Plan t Fire HRA accounts for dependencies in two ways and the results of the visual inspection did not find any additiona l dependencies, this Suggestion level F&O is consider ed closed. However, since the fire response procedures are being updated and trained upon as part of the transition to NFPA 805, it is recognized that the Fire HRA dependency will need to be re-visited during the implementation phase as part of Implementation Item 11-805-090 of Table S-3 in the Callaway Plant NFPA 805 License Amendment Request Attachment S.
 
Completeness uncertainty does apply to fire HRA, as described in the te xt of Section 3.3.1 of the Uncertainty and Sensitivity Analysis (Calculation 17671-014) and implemented in Table 3-
: 1. Table 3-1 documents sources of uncertainty in the Callaway Plant FPRA. Columns 1 through 3 list each NUREG/CR- 6850 FPRA task and summarize the generic treatment of uncertainty issues based on Appendix V of NUREG/CR-6850. Appendix V segregates these issues as either relating to uncertainty, or relating to accuracy and completeness. Since the Fire HRA task uses other NUREG/CR-6850 tasks such as component selection, plant response model, and fire modeling as inputs then the completeness of these inputs affects the completeness of the Fire HRA.
Additionally during the Fi re Risk Evaluation process uncertainty was considered as described in Section 3.7 and as documented in Table 4-3 and Table 4-5 of each Fire Risk Evaluation report (Calculation 17671-FRE-X-YY, where X-YY is the individual fire area analyzed). Table 4-3 documented the importance of applicable re covery actions by showing the risk increase if the recovery action failure pr obability was increased. As a future task during the implementation phase, the treatment of uncertainty related to the Fire HRA will be updated as part of implementation item number 11-805-090 of Table S-3 in the Callaway Plant NFPA 805 License Amendment Request.
 
f) The LERF model and the core damage model were developed in paralle l and have both been available for sensitivity studies since the Oc tober 2009 peer review. All sensitivity studies report risk results for CDF, LERF, CDF, and LERF. Sensitivity studies were performed for two considerations. to ULNRC-05876
 
Page 80 of 136
: 1) Sensitivity studies were performed during the Fire Risk Evaluations to determine effectiveness of a recovery action. These are documented in the respective FRE reports.
: 2) Sensitivities were done on the global core dama ge and LERF equations as part of Task 14.
The Callaway Plant Fire PRA reported two sensitivity studies, which are documented in Appendix B of Calculation 17671-014, "Uncertainty and Sensitivity Analyses". One sensitivity study was the FAQ 08-0048 requirement to us e NUREG/CR-6850 ignition frequencies for ignition bins which are described by a gamma distribution which has an alpha factor of less than 1.0. The other sensitivity study considered the potential risk impact of small quantities of thermoplastic cable, which were discovered in limited locations in the plant. Both of these sensitivity studies considered core damage and LERF metrics, the results of which are reproduced below.
There are two sensitivity studies discussed in this RAI response: the FAQ 08-0048 ignition frequency study and the thermoplastic cable sensitivity. The FAQ 08-0048 sensitivity study results are shown in Table 1, which is a copy of Table B.1-2 of re port 17671-014. This shows the total for each risk metric for fire scenarios that have ignition sources from the applicable bins, AND that contribute a non-zero delta risk to the plan t-wide results.
The thermoplastic sensitivity looked at two different effects of the non-IEEE-383 cable. The first effect was the reduction of damage threshold temperature for the cables as fire targets.
Essentially, this effect increased the target set for certain fi res because the non-IEEE-383 cable fails at a lower temperature, so the ZOI with respect to those cables is larger. The effect of modifying the ZOI to account for these cables is shown in Tables 2 and 3, which correspond to Tables B.2-1 and B.2-2, respectively, in the 17671-014 report.
 
As shown in Table 2, there was no change in risk for the affected fixed ignition sources. The affected transient sources showed a small risk increase. The last portion of the non-IEEE-383 cable sensitivity assigned the full self-ignited cab le bin (bin 12) ignition frequency to each of the affected fire scenarios indi vidually, and the scenario with the largest risk increases was selected as the bounding example. Since the fu ll bin 12 ignition frequency was added to each scenario individually, the results should be l ooked at independently for each scenario, not summed. The results are shown in Table 4, with the largest risk increase being attributed to scenarios 3402-T1 and 3502-T2.
 
As shown in Table 4, the bounding risk increase due to the potential for self-ignited cable fires in non-IEEE-383 rated cable is 8.78E-7/yr. for fi re CDF and 2.28E-8/yr. for fire LERF. There is no potential increase in delta risk because the VFDRs in the Turbine Building are not modeled in the fire PRA. These risk metric in creases are very conservative because the entire bin 12 frequency was applied to each single fire scenario. A more realistic treatment would split the bin 12 frequency amongst the various fire scenarios based on cable weighting factors.
 
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Page 81 of 136 Table 1: FAQ 08-0048 IEF Sensitivity Results Summary
[B.1-2 in 17671-014] Risk Metric Baseline IEF (EPRI) Adjusted (NUREG 6850) IEFDelta Plant Totals (Baseline) CDF (/yr) 3.17E-06 6.81E-06 3.64E-062.02E-05 LERF (/yr) 6.39E-08 1.42E-07 7.77E-083.97E-07 CDF (/yr) 1.31E-06 3.26E-06 1.95E-061.87E-06 LERF (/yr)2.78E-08 7.28E-08 4.50E-083.84E-08 to ULNRC-05876 Page 82 of 136
 
Table 2: Fixed Sources with Modified ZOI
[B.2-1 in 17671-014]
Affected Fire Scenario Ign Source Final IEF (/yr) Baseline Thermoplastic Delta CCDP CDF
(/yr) CLERP LERF 
(/yr) CCDP CDF
(/yr.) CLERP LERF
(/yr.) CDF
(/yr.) LERF
(/yr.) 4316-1 HF187A 2.79E-05 6.14E-04 1.71E-08 1.49E-05 4.16E-10 6.14E-04 1.71E-08 1.49E-05 4.16E-10 0.00E+00 0.00E+00 4316-2 HF187B 2.79E-05 6.14E-04 1.71E-08 1.49E-05 4.16E-10 6.14E-04 1.71E-08 1.49E-05 4.16E-10 0.00E+00 0.00E+00
 
Table 3: Transient Sources with Modified ZOI
[B.2-2 in 17671-014] Affected Fire Scenario Base IEF (/yr.) Baseline Thermoplastic CCDP CLERP Baseline Thermoplastic Delta Wg*SF*Pns Final IEF Wg*SF*Pns Final IEF CDF (/yr.) LERF (/yr.) CDF (/yr.) LERF (/yr.) CDF (/yr.) LERF (/yr.) 4316-T1 3.25E-03 1.44E-03 4.67E-06 1.95E-03 6.34E-06 6.15E-04 1.49E-05 2.87E-09 6.97E-11 3.90E-09 9.46E-11 1.03E-09 2.49E-11 4316-T3 3.25E-03 1.17E-01 3.79E-04 1.16E-01 3.77E-04 4.49E-04 9.93E-06 1.70E-07 3.76E-09 1.70E-07 3.75E-09 -8.55E-10 -1.89E-11 3402-T1 2.44E-04 2.33E-02 5.67E-06 6.46E-02 1.58E-05 6.65E-04 1.73E-05 3.77E-09 9.79E-11 1.05E-08 2.72E-10 6.71E-09 1.74E-10 3402-T2 2.44E-04 9.77E-01 2.38E-04 9.35E-01 2.28E-04 6.15E-04 1.49E-05 1.46E-07 3.55E-09 1.40E-07 3.40E-09 -6.26E-09 -1.52E-10 3502-T2 3.56E-04 2.18E-02 7.78E-06 3.40E-02 1.21E-05 6.65E-04 1.73E-05 5.18E-09 1.34E-10 8.06E-09 2.09E-10 2.88E-09 7.47E-11 3502-T4 3.56E-04 7.02E-02 2.50E-05 5.81E-02 2.07E-05 7.87E-05 6.68E-08 1.97E-09 1.67E-12 1.63E-09 1.38E-12 -3.38E-10 -2.87E-13 Totals 3.31E-07 7.62E-09 3.34E-07 7.72E-09 3.16E-09 1.03E-10 to ULNRC-05876
 
Page 83 of 136 Table 4: Bin 12 IEF Scoping Analysis
[B.2-3 in 17671-014]
Affected Fire Scenario Baseline IEF  (per yr.)
FAQ 0048 Bin 12 IEF (/yr.) CCDP CLERPBaseline Bin 12 Delta Bin 12 Freq. (/yr.) CDF (per yr.)
LERF (per yr.) CDF (per yr.) LERF (per yr.) CDF (per yr.)
LERF (per yr.) 4316-1 2.79E-05 1.32E-03 1.35E-03 6.14E-04 1.49E-05 1.71E-08 4.16E-10 8.28E-07 2.01E-08 8.11E-07 1.97E-08 4316-2 2.79E-05 1.32E-03 1.35E-03 6.14E-04 1.49E-05 1.71E-08 4.16E-10 8.28E-07 2.01E-08 8.11E-07 1.97E-08 4316-T1 4.67E-06 1.32E-03 1.32E-03 6.15E-04 1.49E-05 2.87E-09 6.97E-11 8.14E-07 1.98E-08 8.11E-07 1.97E-08 4316-T3 3.79E-04 1.32E-03 1.70E-03 4.49E-04 9.93E-06 1.70E-07 3.76E-09 7.63E-07 1.69E-08 5.93E-07 1.31E-08 3402-T1 5.67E-06 1.32E-03 1.33E-03 6.65E-04 1.73E-05 3.77E-09 9.79E-11 8.82E-07 2.29E-08 8.78E-07 2.28E-08 3402-T2 2.38E-04 1.32E-03 1.56E-03 6.15E-04 1.49E-05 1.46E-07 3.55E-09 9.58E-07 2.32E-08 8.11E-07 1.97E-08 3502-T2 7.78E-06 1.32E-03 1.33E-03 6.65E-04 1.73E-05 5.18E-09 1.34E-10 8.83E-07 2.29E-08 8.78E-07 2.28E-08 3502-T4 2.50E-05 1.32E-03 1.34E-03 7.87E-05 6.68E-08 1.97E-09 1.67E-12 1.06E-07 8.99E-11 1.04E-07 8.82E-11
 
to ULNRC-05876 Page 84 of 136 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 Me thodology 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 admini strative controls is considered to be already embedded in the transient fire frequencies based on the historical data. Furthermore, a physical analysis unit with a tota l 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 th an 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 Calcula tion, states that (1) the Callaw ay 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. Nonethel ess, 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 pl ants 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 a lternate approach to justify the use of such a factor.
 
Additional Justification Needed:
Include in the results the quantitative results, if any, on both delta-CDF and delta-LERF as well, since YD-1 has an associated VFDR.  (Note: It appears that the answer is no change, since the VFDR does not appear to be related to the bus ducts, but this needs to be docketed to complete the response).
Response to Probabilistic Risk Assessment RAI-08
: a. Response provided by ULNRC-05851 dated April 17, 2012.
: b. Ignition bin 16.2 in NUREG
/CR-6850, Supplement 1 involves iso-phase bus ducts. Callaway Plant uses considerably fewer iso-phase bus ducts than a typical plant. Callaway Plant preferentially uses    to ULNRC-05876
 
Page 85 of 136 cable ducts for termination of high energy distribution points. In accordance with FAQ-0035, fires
 
in cable ducts have been incorporated with the end device, thus they need no specific initiating event treatment. All of the Bin 16.2 components at Callaway Plant are contained within the fire area YD-1 Yard, specifically the Circulating Water Pump House (CWPH). Callaway Plant has 9 bus ducts, whereas the typical plant has approximately 45 bus duct components.
A sensitivity study was performed to quantify the risk impact of using the full bin 16.2 ignition frequency in the Callaway Plant FPRA. This sensitivity removes the factor of 5 reduction (i.e., 5 =
 
45/9) in the bin which was used in the Initiating Event Frequenc y Calculations (Calculation 17671-005) and recalculates the fire risk in the affected Fire Area(s).
Table 1 shows the results of this sensitivity study. The line for "LAR Values" reflects the base case ignition source values in the CWPH area. The line for "Full Bin 16.2" case reflects the full value for the bin 16.2 from NUREG/CR-6850 (Supplement 1). The results for each case, and the change in frequencies, are shown in the table below.
 
Table 1 - Bin 16.2 Full Value Sensitivity Results Fire area YD-1 Yard has a total Fire CDF of 1.03E-6/yr. and a total Fire LERF of 2.18E-8/yr. in the Callaway Plant NFPA 805 LAR submittal. Compared to the LAR risk totals, the CDF and LERF increases shown in Table 1 are less than 1% of each metric.
 
Fire Area YD-1 has a VFDR due to cables for the Refueling Water Storage Tank (RWST) water level sensors. These cables are not present in the Circulating Water Pumphouse, where all of the bus ducts are located. There are no VFDRs in the Circulating Water Pump House Area. There is no change in delta risk due to an increase in the IE freque ncy of bin 16.2. The increase in deterministic risk due to using the full bin 16.2 ignition frequency is considered to be negligible.
Case IEF  (per yr.) CCDP CLERP CDF  (per yr.)
LERF (per yr.) CDF  (per yr.) LERF (per yr.)
LAR Values 5.18E-03 1.24E-05 2.41E-07 6.42E-08 1.25E-09 0.00E+00 0.00E+00 Full Bin 16.2 5.84E-03 1.24E-05 2.41E-07 7.24E-08 1.41E-09 0.00E+00 0.00E+00 Frequency Increase 6.60E-04 n/a n/a 8.18E-09 1.59E-10 0.00E+00 0.00E+00    to ULNRC-05876
 
Page 86 of 136 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 Iden tification and Ranking Table (PIR T) 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 estimati ng 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 fa ilure (or spurious operation) was due to fire-induced cable impacts from the subsequent analyses. Note also that assuming the presence of CPTs for control circuits in the MCR panels may be incorrect and, if so, should be removed when performing the sensitivity analysis.
Additional Justification Needed
 
At the end of the response, the response to PRA RAI 13 is cited as a basis for concluding that the results in this RAI (9a) are bounding and conser vative due to failure probabilities exceeding one. In the response to PRA RAI 13, was not the evaluation performed such that the effect of probabilities exceeding one was eliminated?  Also, are all the effects from the RAI 13 Response included in the changes made in performing the sensitivity analysis for RAI 9a?
: b. The Uncertainty and Sensitivity Analyses Calculation indicates that sensitivity/uncertainty analyses were not performed for fire ignition frequencies (other than the bins required by FAQ 48 in Supp. 1 to NUREG/CR-6850) or cable failure mode likelihoods. Please provide the results of sensitivity/uncertainty analyses for these values.
 
Additional Justification Needed When combining the risk and delta-risk incr eases per bin in Table 4, does the following correctly characterize how the w hole-area burnup scenario contributions were included?  For all the areas assumed to contribute per a sing le bin, e.g., A-28, et al
., to bin 15.1, or grouped bin, e.g., A-3, et al., to grouped bin 5/6/7 in Table 3, the contribution from that area was included with the corresponding bin (or grouped bin) in Table 4. That is, while for scenarios with a single ignition source th e contribution arose solely fr om the actual corresponding bin, for the whole-area burnup scenarios the contribution arose solely from the assumed corresponding bin, as per Table 3. If this is not the correct characterization, provide what is.
to ULNRC-05876
 
Page 87 of 136 Response to Probabilistic Risk Assessment RAI-09
: a. The Callaway Plant FPRA did not use low spur ious actuation probabilities to screen out components from inclusion in the FPRA. As such, the potentially increased spurious probabilities due to elimination of credit for control power transformers (CPT) have no effect on what components are included in the FPRA.
 
A sensitivity study has been performed to examine the effect of removing credit for CPTs in the calculation of spurious failure probabilities for AC powered control circuits. Merged, global risk equations were used, with appropriate manipulation of spurious failure probabilities, to estimate the impact on fire CDF, CDF, LERF, and LERF.
The impact on total CDF and total LERF is a straig ht forward process. All fire scenarios in the PRA are concatenated into a single equation and the failure probabilities are manipulated to increase the probabilities of certain basic events. 
 
The impact on delta CDF/LERF is more difficult to calculate, so a surrogate measure is used. Delta risk is the variant risk minus the compliant risk. Both variant risk and compliant risk can be represented by a cutset equation, which can be quantified to find the equation value. Delta risk is the difference between the variant risk value and compliant risk value. There is no cutset equation for delta risk. Delta risk is calculated during the Fire Risk Evaluation (FRE) process for each variant scenario. The total plant delta risk is the addition of delta risk from all variant scenarios. In order to calculate the delta risk exactly in this sensitivity study, it would be necessary to re-quantify each FRE. 
 
As an alternative approach, all variant scenar ios were concatenated into an equation and quantified. These merged, "VFDR-only" equations were modified with the appropriate increased spurious event probabilities and the increase in risk was assumed to be equal to the increase in delta risk. The cha nge in risk for this equation wa s calculated and taken to be the maximum possible change in delta risk due to eliminating credit for CPTs. 
 
The first step was to identify all AC powered components that credit a CPT for the calculation of spurious failure probabilities. Table 1 shows these components.
 
Table 1 Components with CPT Credit Equipment ID Component Type ALHV0005 MOV ALHV0007 MOV ALHV0009 MOV ALHV0011 MOV ALHV0030 MOV ALHV0031 MOV ALHV0032 MOV ALHV0033 MOV    to ULNRC-05876
 
Page 88 of 136 Table 1 Components with CPT Credit Equipment ID Component Type ALHV0034 MOV ALHV0035 MOV ALHV0036 MOV BBHV0013 MOV BBHV0014 MOV BBHV0015 MOV BBHV0016 MOV BBHV8000A MOV BBHV8000B MOV BBHV8351A MOV BBHV8351B MOV BBHV8351C MOV BBHV8351D MOV BBPV8702A MOV BBPV8702B MOV BGHV8105 MOV BGHV8105 MOV BGHV8106 MOV BGHV8106 MOV BGHV8109 MOV BGHV8110 MOV BGHV8110 MOV BGHV8111 MOV BGHV8111 MOV BGHV8357A MOV BGHV8357B MOV BGLCV0112B MOV BGLCV0112C MOV BNHV8806A MOV BNHV8806B MOV BNHV8812A MOV BNHV8812B MOV BNHV8813 MOV BNLCV0112D MOV BNLCV0112E MOV EAHV0005 HDV EAHV0006 HDV EFHV0037 MOV    to ULNRC-05876
 
Page 89 of 136 Table 1 Components with CPT Credit Equipment ID Component Type EFHV0038 MOV EFHV0051 MOV EFHV0052 MOV EFHV0059 MOV EFHV0060 MOV EFHV0066 MOV EGHV0011 MOV EGHV0012 MOV EGHV0013 MOV EGHV0014 MOV EGHV0015 MOV EGHV0015 MOV EGHV0016 MOV EGHV0016 MOV EGHV0053 MOV EGHV0053 MOV EGHV0054 MOV EGHV0054 MOV EGHV0058 MOV EGHV0061 MOV EGHV0062 MOV EGHV0069A AOV EGHV0069B AOV EGHV0070A AOV EGHV0070B AOV EGHV0071 MOV EJFCV0610 MOV EJFCV0611 MOV EJHV8701A MOV EJHV8701B MOV EJHV8716A MOV EJHV8716B MOV EJHV8804A MOV EJHV8804B MOV EJHV8809A MOV EJHV8809B MOV EJHV8811A MOV EJHV8811B MOV    to ULNRC-05876
 
Page 90 of 136 Table 1 Components with CPT Credit Equipment ID Component Type EMHV8801A MOV EMHV8801B MOV EMHV8803A MOV EMHV8803B MOV EMHV8814A MOV EMHV8814B MOV EMHV8821A MOV EMHV8821B MOV EMHV8923A MOV EMHV8923B MOV ENHV0001 MOV ENHV0006 MOV ENHV0007 MOV ENHV0012 MOV EPHV8808A MOV EPHV8808B MOV EPHV8808C MOV EPHV8808D MOV LFFV0095 MOV NB0109 BKR NB0112 BKR NB0209 BKR NB0212 BKR PA0201 BKR VEA2101A HDV VEA2101B HDV Where:  MOV = Motor Operated Valve
 
AOV = Air Operated Valve
 
BKR = Breaker
 
HDV = Hydraulically Driven Valve
 
The next step was to identify base case spurious failure probabilities associated with each component in Table 1. Since this sensitivity uses the global results equations for each risk
 
metric, this list of basic events was trimmed down by only including spurious failure basic events that actually appear in at least one of the results equations. Table 2 shows:
a) all components appearing in the fi nal risk equation (CDF or LERF) b) basic event name(s) for these components c) the base case value for spuri ous actuation of these com ponents which credits the CPT    to ULNRC-05876
 
Page 91 of 136 d) the sensitivity study value for spurious operation, which does not credit CPT.
Table 2 - Modified Spuri ous Failure Basic Events Equipment ID Spurious BE(s)
LAR Value PRA RAI 09a Probability ALHV0034 AL-MOV-SC-ALHV34 0.4 0.8 ALHV0035 AL-MOV-SC-ALHV35 0.4 0.8 ALHV0036 AL-MOV-SC-ALHV36 0.4 0.8 BBHV0013 BB-MOV-SC-HV0013 0.4 0.8 BB-MOV-SC-HV00133:00E-01 0.3 0.6 BBHV0014 BB-MOV-SC-HV0014 0.4 0.8 BB-MOV-SC-HV00143:00E-01 0.3 0.6 BBHV0015 BB-MOV-SC-HV0015 0.4 0.8 BB-MOV-SC-HV00153:00E-01 0.3 0.6 BBHV0016 BB-MOV-SC-HV0016 0.4 0.8 BB-MOV-SC-HV00163:00E-01 0.3 0.6 BBHV8000A BB-MOV-SC-8000A 0.4 0.8 BBHV8000B BB-MOV-SC-8000B 0.4 0.8 BBHV8351A BB-MOV-SC-V8351A 0.4 0.4* BB-MOV-SC-V8351A3:00E-01 0.3 0.3*
BBHV8351B BB-MOV-SC-V8351B 0.4 0.4* BB-MOV-SC-V8351B3:00E-01 0.3 0.3*
BBHV8351C BB-MOV-SC-V8351C 0.4 0.4* BB-MOV-SC-V8351C3:00E-01 0.3 0.3*
BBHV8351D BB-MOV-SC-V8351D 0.4 0.4* BB-MOV-SC-V8351D3:00E-01 0.3 0.3*
BGLCV0112B BG-MOV-SC-V112B 0.4 0.8 BG-MOV-SC-V112B1:00E-01 0.1 0.2 BGLCV0112C BG-MOV-SC-V112C 0.4 0.8 BNHV8806A BN-MOV-SC-V8806A 0.4 0.8 BNHV8806B BN-MOV-SC-V8806B 0.4 0.8 BNHV8813 BN-MOV-SC-HV8813 0.4 0.8 EAHV0005 EA-HDV-SC-HV0005 0.4 0.8 EA-HDV-SC-HV00053:00E-01 0.3 0.6 EAHV0006 EA-HDV-SC-HV0006 0.4 0.8 EA-HDV-SC-HV00063:00E-01 0.3 0.6 EFHV0051 EF-MOV-SC-EFHV51 0.4 0.8 EFHV0052 EF-MOV-SC-EFHV52 0.4 0.8 EFHV0059 EF-MOV-SO-EFHV59 0.4 0.8 EF-MOV-SO-EFHV593:00E-01 0.3 0.6 EFHV0060 EF-MOV-SO-EFHV60 0.4 0.8    to ULNRC-05876
 
Page 92 of 136 Table 2 - Modified Spuri ous Failure Basic Events Equipment ID Spurious BE(s)
LAR Value PRA RAI 09a Probability EF-MOV-SO-EFHV603:00E-01 0.3 0.6 EGHV0011 EG-MOV-SO-HV11 0.4 0.8 EGHV0012 EG-MOV-SO-HV12 0.4 0.8 EGHV0013 EG-MOV-SO-HV13 0.4 0.8 EGHV0014 EG-MOV-SO-HV14 0.4 0.8 EGHV0058 EG-MOV-SC-EGHV58 0.4 0.8 EG-MOV-SC-EGHV583:00E-01 0.3 0.6 EGHV0061 EG-MOV-SC-EGHV61 0.4 0.8 EG-MOV-SC-EGHV613:00E-01 0.3 0.6 EGHV0062 EG-MOV-SC-EGHV62 0.4 0.8 EG-MOV-SC-EGHV623:00E-01 0.3 0.6 EGHV0071 EG-MOV-SC-HV00713:00E-01 0.3 0.6 EG-MOV-SC-HV00714:00E-01 0.4 0.8 EJFCV0611 EJ-MOV-SC-FCV611 0.4 0.8 EJHV8716A EJ-MOV-SC-V8716A 0.4 0.8 EJHV8716B EJ-MOV-SC-V8716B 0.4 0.8 EJHV8811A EJ-MOV-SO-V8811A 0.4 0.8 EJ-MOV-SO-V8811A3:00E-01 0.3 0.6 EJHV8811B EJ-MOV-SO-V8811B 0.4 0.8 EJ-MOV-SO-V8811B3:00E-01 0.3 0.6 EMHV8801A EM-MOV-SO-V8801A 0.4 0.8 EM-MOV-SO-V8801A8:00E-01 0.8 0.8 EMHV8801B EM-MOV-SO-V8801B 0.4 0.8 EM-MOV-SO-V8801B8:00E-01 0.8 0.8 EMHV8803A EM-MOV-SO-V8803A 0.4 0.8 EMHV8803B EM-MOV-SO-V8803B 0.4 0.8 EMHV8814B-
 
PRA EM-MOV-SC-V8814B 0.4 0.8 EM-MOV-SC-V8814B8:00E-01 0.8 0.8 EMHV8923A EM-MOV-SC-V8923A 0.4 0.8 ENHV0001 EN-MOV-SO-ENHV01 0.4 0.8 ENHV0006 EN-MOV-SO-HV0006 0.4 0.8 EN-MOV-SO-HV00063:00E-01 0.3 0.6 ENHV0007 EN-MOV-SO-ENHV07 0.4 0.8 ENHV0012 EN-MOV-SO-HV0012 0.4 0.8 EN-MOV-SO-HV00123:00E-01 0.3 0.6 EPHV8808A EP-MOV-SC-V8808A 0.4 0.8 EPHV8808B EP-MOV-SC-V8808B 0.4 0.8 EPHV8808C EP-MOV-SC-V8808C 0.4 0.8    to ULNRC-05876
 
Page 93 of 136 Table 2 - Modified Spuri ous Failure Basic Events Equipment ID Spurious BE(s)
LAR Value PRA RAI 09a Probability EPHV8808D EP-MOV-SC-V8808D 0.4 0.8 NB0212 NB-BKR-SC-NB0212 0.1 0.2 PA0201 PA-BKR-SO-PA0201 0.4 0.8
*No change. See explanation below for why these probabilities do not require change.
Explanation for BBHV8351A-D:
The probabilities for spurious closure of BBHV8 351A, B, C and D were not increased in the sensitivity study, even though the base case values credit a CPT. The probabilities in the base case are high enough that they violate the rare event approximati on for fault tree logic codes. These four valves appear in an 'OR' gate for loss of seal injection flow. These four valves always appear in the same scenario. There are no scenarios with one or two valves. The current probability for loss of seal cooling when these basic events appear in a scenario is greater than
: 1. Increasing the probabilities for these valv es would only increase the over-counting for loss of seal cooling. 
 
There are two sets of probabilities that maximize seal cooling loss at a value of 1.0. These are:
a) each valve is assigned a .25, or b) one valve is assigned a 1.0 and the other 3 are assigned a 0.0.
For the sensitivity study, the valv e probabilities for BBHV8351A-D were retained at the base case values of 0.3 and 0.4, which resu lts in a total scenario probabi lity for loss of seal cooling of 1.2 to 1.6.
 
As seen in Table 2, some events have multiple probabilities used for the same event. This occurs because in some instances, the component is onl y susceptible to inte rnal hot shorts and in some cases the cable is also susceptible to ex ternal hot shorts. The 0.3 was used for fires in the main control board where external hot shorts were not considered valid. For the sensitivity study, all the base case (LAR) probabilities are doubled to cr eate the value used in the sensitivity study. 
 
The updated spurious failure probabilities were then imported into the global basic event data (BED) files for each of the four (CDF, CDF, LERF, and LERF) global risk equations. The resulting increase in each metr ic is shown in Table 3.
to ULNRC-05876
 
Page 94 of 136 Table 3 - Risk Metric Increases Case Description Risk Increase CDF (/yr) Estimate of plant-wide CDF increase 6.26E-06 LERF (/yr) Estimate of plant-wide LERF increase 1.48E-07 CDF-VFDR (/yr) Estimate of plant-wide CDF increase 1.75E-06 LERF-VFDR (/yr) Estimate of plant-wide LERF increase 7.36E-08
 
Table 4 provides some perspective by comparing these risk metric increases to their corresponding plant-wide totals reported in the license amendment request, and comparing the theoretical total of the tw o against the risk goals from Regulatory Guide (RG) 1.205.
Table 4 - Comparison of Risk Increases to Risk Goals Metric Increase LAR Plant Total Theoretical Total RG 1.205 Goal CDF (/yr) 6.26E-06 2.04E-05 2.67E-05 < 1E-4/yr. LERF (/yr) 1.48E-07 3.97E-07 5.45E-07 < 1E-5/yr. CDF (/yr) 1.75E-06 1.87E-06 3.62E-06 < 1E-5/yr. LERF (/yr) 7.36E-08 3.84E-08 1.12E-07 < 1E-6/yr.
 
As shown, when the risk increases due to de-crediting the CPTs are added to the existing baseline risk metrics, the resultant theoretical totals are still below the risk goals presented in RG 1.205. In addition, increasing the probability of spurious operation to 0.8 for all MOV's and AOV's causes several PRA functions (such as loss of all RCP seal cooling, loss of CST inventory, loss of Steam Generator cooling) to become signifi cantly greater than 1.0. This issue is discussed specifically for the BBHV 8351A-D valves under Table 2 above and is also discussed in PRA RAI-13 to explain the generation of negative risk numbers. The issue involves using probabilities for basic events in an 'OR' gate that vi olate the "rare event approximation", which is required by the WINNUPRA code for representative results. This issue has been identified and isolated to certain scenarios for the seal injection valves (BBHV8351A-D) and the CST drain valves (ADLV0079BA/BB), a nd is therefore possible to fix the issue with global PRA data changes. The issue is known to occur for other PRA functions, but it is not possible to isolate the issue to certain scenarios, so it is not possible to make global data changes to correct the issu
: e. The over counting issue for the BBHV8351A-D valves was corrected for this sensitivity study.
The over counting issue was not corrected for the CST drain scenario (valves ADLV0079BA/BB), nor any other PRA function. As such, the risk numbers presented in this sensitiv ity are considered c onservative and bounding.
 
The issues from the PRA RAI-13 response (i.e., PO RV and block valve op erability) were not incorporated into this RAI sensitivity study.
However, the effects of CPT credit (including retaining the BBHV8351A-D probabi lities at the base case 0.4) were incorporated into PRA RAI-13.
to ULNRC-05876
 
Page 95 of 136
: b. A sensitivity or uncertainty study on the cable failure likelihoods, beyond what was performed
 
in part A of this RAI response, is not considered necessary. In part A of this RAI response, removing Control Power Transformer (CPT) credit in the cable failure likelihood calculations caused the vast majority of spurious failures in the Callaway Plant FPR A to have a value of 0.8. With so many values close to 1.0, there is little remaining uncertainty towards the upper bound of the uncertainty distribution, so the CPT de-crediting sensitivity in part A of this RAI response is considered sufficiently bounding. A sensitivity study was performed to consider the effects of ignition frequency uncertainty on fire risk, using the uncertainty parameters presented in NUREG/CR-6850, Supplement 1. The FPR A results, as presented in the Callaway Plant LAR, utilize the mean i gnition frequency values presented in Supplement 1. In this sensitivity study, the ignition frequencies at the 95% confidence interval for each "bin" were
 
used, and the effect on fire CDF, LERF, delta CDF (CDF), and delta LERF (LERF) was measured. The effects on these four risk metrics were calculated for each ignition frequency bin individually, and then all bins were combined to see the effect of using all bins at the 95% confidence interval frequency simultaneously. This second study is more severe than a parametric uncertainty run using a Crystal Ball or a Monte Carlo approach, the final value of the sensitivity study assumes all fire frequencies are simultaneously at their 95% level.
Technical Approach
 
The basis for the study was to multiply all scenario base line CDF/LERF values by a factor representing the multiplication factor for the 95th percentile bin ignition frequency. The first step was to determine the multiplication factor for each bin. This is simply the 95% frequency divided by the mean frequency. The multiplication factors for each bin are shown in Table 1.
Note that a special bin, 33/34/35, is defined for use with the catastrophic Turbine-Generator fire, which is dictated by the me thodology in Appendix O of NUREG/CR-6850.
Table 1 - Ignition Frequency Bin Multipliers Bin No. Location Ignition Source Mean value (used in base case quant.) 95 th percentile bin ignition frequency Multiplier used for study 1 Battery Room Batteries 3.26E-04 1.25E-03 3.83 2 Containment (PWR) Reactor Coolant Pump 2.35E-03 6.11E-03 2.60 3 Containment (PWR)
Transients and hot
 
work 2.34E-03 5.89E-03 2.52 4 Control Room Main Control
 
Board 8.24E-04 2.47E-03 3.00 5 Control/Aux/Reactor Building Cable fires caused
 
by welding and cutting 1.25E-03 2.83E-03 2.26    to ULNRC-05876
 
Page 96 of 136 Table 1 - Ignition Frequency Bin Multipliers Bin No. Location Ignition Source Mean value (used in base case quant.) 95 th percentile bin ignition frequency Multiplier used for study 6 Control/Aux/Reactor Building Transient fires
 
caused by welding and cutting 2.46E-03 5.65E-03 2.30 7 Control/Aux/Reactor Building Transients 4.81E-03 8.80E-03 1.83 8 Diesel Generator Room Diesel Generators 5.04E-03 9.02E-03 1.79 9 Plant-Wide Components Air Compressors 4.65E-03 8.51E-03 1.83 10 Plant-Wide Components Battery Chargers 1.18E-03 3.07E-03 2.60 11 Plant-Wide Components Cable fires caused
 
by welding and cutting 9.43E-04 2.82E-03 2.99 12 Plant-Wide Components Cable Run (Self-
 
ignited cable fires) 1.32E-03 3.43E-03 2.60 13 Plant-Wide Components Dryers 4.20E-04 1.61E-03 3.83 14 Plant-Wide Components Electric motors 3.41E-03 6.61E-03 1.94 15.1 Plant-Wide Components Electrical Cabinets Non-HEAF 2.36E-02 9.40E-02 3.98 15.2 Plant-Wide Components Electrical
 
Cabinets-HEAF 1.06E-03 2.75E-03 2.59 16.1 Plant-Wide Components Bus Ducts 1.27E-03 3.31E-03 2.61 16.2 Plant-Wide Components Iso-phase Bus
 
Ducts 1.65E-04 1 2.15E-03 13.03 17 Plant-Wide Components Hydrogen Tanks 1.18E-03 3.07E-03 2.60 18 Plant-Wide Components Junction box 1.11E-03 2.89E-03 2.60 19 Plant-Wide Components Misc. Hydrogen
 
Fires 1.24E-03 3.22E-03 2.60 20 Plant-Wide Components Off-gas/H2 Recombiner (BWR) 8.83E-03 1.95E-02 2.21 21 Plant-Wide Components Pumps 1.42E-02 2.06E-02 1.45    to ULNRC-05876
 
Page 97 of 136 Table 1 - Ignition Frequency Bin Multipliers Bin No. Location Ignition Source Mean value (used in base case quant.) 95 th percentile bin ignition frequency Multiplier used for study 22 Plant-Wide Components RPS MG sets 9.33E-04 2.88E-03 3.09 23 Plant-Wide Components Transformers 8.02E-03 1.29E-02 1.61 24 Plant-Wide Components Transient fires
 
caused by welding and cutting 3.65E-03 7.38E-03 2.02 25 Plant-Wide Components Transients 8.28E-03 1.37E-02 1.65 26 Plant-Wide Components Ventilation Subsystems 6.12E-03 1.04E-02 1.70 27 Transformer Yard Transformer -
 
Catastrophic 1.62E-03 4.21E-03 2.60 28 Transformer Yard Transformer - Non
 
Catastrophic 8.38E-03 1.40E-02 1.67 29 Transformer Yard Yard transformers (Others) 1.89E-03 3.79E-03 2.01 30 Turbine Building Boiler 9.78E-04 2.55E-03 2.61 31 Turbine Building Cable fires caused
 
by welding and cutting 4.50E-04 1.73E-03 3.84 32 Turbine Building Main feedwater pumps 5.44E-03 1.00E-02 1.84 33 Turbine Building Turbine Generator (T/G) Exciter 2.10E-03 4.98E-03 2.37 34 Turbine Building T/G Hydrogen 3.23E-03 6.79E-03 2.10 35 Turbine Building T/G Oil 3.89E-03 7.82E-03 2.01 36 Turbine Building Transient fires
 
caused by welding and cutting 7.55E-03 1.28E-02 1.70 37 Turbine Building Transients 3.41E-03 7.07E-03 2.07 33/34/35 Turbine Building Catastrophic TG
 
Fire 9.22E-03 1.96E-02 2.12
 
Note 1 - The listed mean frequency for bin 16.2 is the NUREG/CR-6850 value reduced by a factor of 9/45 to account for the limited use of iso-phase bus ducts at Callaway Plant. This is the value used in the Callaway Plant LAR results. The 95 th value is the full frequency from NUREG/CR-6850, Supplement 1. to ULNRC-05876
 
Page 98 of 136 For transient fires at Callaway Plant, the room ignition frequencies are comprised of contributions from multiple bins. To simplify this sensitivity study and provide a conservative answer, the maximum multiplier from all transient bins for that area was used as a multiplier for the entire area fire
 
transient initiating frequency. Bin 3 is an exception since transient fires in Containment have their own special bin. The multipliers used for transient fires are shown in Table 2.
 
Table 2 - Multipliers for Transient Ignition Frequencies Transient Bins Location Multiplier 5/6/7 Control/Aux/Reactor Building 2.30 31/36/37 Turbine Building 3.84 11/24/25 Plant-Wide Components 2.99 3 Containment (PWR) 2.52
 
Additionally, some fire areas were modeled as whole room burnup. The whole-room burnup scenarios use contributions from multiple ignition frequency bins to create a single, area-wide ignition frequency. To simplify this sensitivity study and provide a conservative answer, the highest single multiplier from all applicable bins for a given fire area were used as a multiplier for the entire area fire ignition frequency. A summary of the "most limiting bin" (MLB) in each w hole-area burnout scenario is shown in Table 3.
 
Table 3 - Most Limiting Bin for Whole-Area Burnup Scenarios Fire Area Scenario MLB MLB Multiplier A-3 A3-WR 5/6/7 2.30 A-5 A5-WR 5/6/7 2.30 A-7 A7-WR 5/6/7 2.30 A-9 A9-WR 5/6/7 2.30 A-10 A10-WR 5/6/7 2.30 A-12 A12-WR 5/6/7 2.30 A-13 A13-WR 5/6/7 2.30 A-14 A14-WR 5/6/7 2.30 A-20 A20-WR 5/6/7 2.30 A-24 A24-WR 5/6/7 2.30 A-25 A25-WR 5/6/7 2.30 A-26 A26-WR 5/6/7 2.30 A-28 A28-WR 15.1 3.98 A-29 A29-WR 5/6/7 2.30 A-30 A30-WR 5/6/7 2.30 A-33 A33-WR 15.1 3.98 AB-1 AB1-WR 15.1 3.98 C-2 C2-WR 5/6/7 2.30 C-3 C3-WR 5/6/7 2.30    to ULNRC-05876
 
Page 99 of 136 Table 3 - Most Limiting Bin for Whole-Area Burnup Scenarios Fire Area Scenario MLB MLB Multiplier C-7 C7-WR 5/6/7 2.30 C-8 C8-WR 5/6/7 2.30 C-13 C13-WR 5/6/7 2.30 C-14 C14-WR 5/6/7 2.30 C-19 C19-WR 5/6/7 2.30 C-20 C20-WR 5/6/7 2.30 C-25 C25-WR 5/6/7 2.30 C-26 C26-WR 5/6/7 2.30 C-28 C28-WR 5/6/7 2.30 C-29 C29-WR 15.1 3.98 C-32 C32-WR 5/6/7 2.30 C-34 C34-WR 5/6/7 2.30 C-35 C35-WR 15.1 3.98 C-36 C36-WR 5/6/7 2.30 C-37 C37-WR 5/6/7 2.30 FB-1 FB-WR 15.1 3.98 LDF-1 LDF1-WR 15.1 3.98 RSB-1 RSB-WR 15.1 3.98 RW-1 RW-WR 15.1 3.98 UNCT UNCT-WR 15.1 3.98 UNPH UNPH-WR 15.1 3.98 USCT USCT-WR 15.1 3.98 USPH USPH-WR 15.1 3.98 YD-1 YD-SWYD 11/24/25 2.99 YD-1 YD-MXFR 11/24/25 2.99 YD-1 YD-SXFR 11/24/25 2.99 YD-1 YD-EX1 11/24/25 2.99 YD-1 YD-EX2 11/24/25 2.99 YD-1 YD-FPH 11/24/25 2.99 YD-1 YD-RWST 11/24/25 2.99 YD-1 YD-CST 11/24/25 2.99 YD-1 YD-CWPH 16.2 13.03 YD-1 YD-UHS 11/24/25 2.99 The CDF, CDF, LERF, and LERF results for each individual and whole-area burnup scenario were then increased with the appropriate multiplier (fixed scenario, transient s cenario, or whole-area burnup scenario). The results are presented by bin, so that the increase in each metric per bin can be seen if the 95% confidence ignition frequency for that particular bin is used. Fire CDF and CDF results are shown in Table 4, and fire LERF and LERF are shown in Table 5. For the areas which only have a    to ULNRC-05876
 
Page 100 of 136 single bin ignition frequency, or a grouped transient bin, from Table 3, the contribution from that area was included with the correspondin g bin (or grouped bin) in Tabl e 4. The transient bins and the catastrophic TG fire are not listed individually, but are listed as a group at the end of the table.
Table 4 - Fire CDF Results by Ignition Frequency Bin Bin # Ignition Source LAR CDF RAI 9b CDF CDF Increase (/yr) LAR CDF (/yr) RAI 9b CDF (/yr) CDF Increase (/yr) 1 Batteries 6.09E-08 2.33E-07 1.72E-07 0.00E+000.00E+00 0.00E+00 2 Reactor Coolant Pump 4.19E-08 1.09E-07 6.70E-08 2.82E-10 7.33E-10 4.51E-10 3 Transients
 
and hot work 1.29E-07 3.24E-07 1.95E-07 1.54E-08 3.88E-08 2.34E-08 4 Main Control
 
Board 6.64E-07 1.99E-06 1.33E-06 6.64E-07 1.99E-06 1.33E-06 8 Diesel Generators 1.14E-08 2.03E-08 8.96E-09 0.00E+000.00E+00 0.00E+00 9 Air Compressors 2.53E-09 4.63E-09 2.10E-09 0.00E+000.00E+00 0.00E+00 10 Battery Chargers 2.37E-07 6.16E-07 3.79E-07 0.00E+000.00E+00 0.00E+00 12 Cable Run (Self-ignited cable fires) 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00E+00 0.00E+0013 Dryers 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00E+00 0.00E+00 14 Electric motors 1.26E-08 2.45E-08 1.18E-08 1.64E-09 3.18E-09 1.54E-09 15.1 Electrical
 
Cabinets Non-HEAF 9.78E-06 3.90E-05 2.92E-05 6.30E-07 2.51E-06 1.88E-06 15.2 Electrical
 
Cabinets-HEAF 2.63E-07 6.83E-07 4.20E-07 3.66E-08 9.49E-08 5.83E-08 16.1 Bus Ducts 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00E+00 0.00E+00 16.2 Iso-phase
 
Bus Ducts 6.42E-08 8.36E-07 7.72E-07 0.00E+000.00E+00 0.00E+00 17 Hydrogen Tanks 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00E+00 0.00E+0018 Junction box 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00E+00 0.00E+00 19 Misc.
Hydrogen Fires 2.79E-09 7.25E-09 4.46E-09 2.79E-09 7.25E-09 4.46E-09    to ULNRC-05876
 
Page 101 of 136 Table 4 - Fire CDF Results by Ignition Frequency Bin Bin # Ignition Source LAR CDF RAI 9b CDF CDF Increase (/yr) LAR CDF (/yr) RAI 9b CDF (/yr) CDF Increase (/yr) 20 Off-gas/ H2 Recombiner (BWR) 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00E+00 0.00E+0021 Pumps 1.29E-06 1.87E-06 5.81E-07 1.66E-08 2.41E-08 7.47E-09 22 RPS MG sets 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00E+00 0.00E+0023 Transformers  2.64E-08 4.25E-08 1.61E-08 5.32E-09 8.56E-09 3.24E-09 27 Transformer
-Catastrophic 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00E+00 0.00E+00 28 Transformer
- Non Catastrophic 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00E+00 0.00E+00 29 Yard transformers (Others) 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00E+00 0.00E+00 26 Ventilation Subsystems 2.11E-07 3.58E-07 1.47E-07 5.25E-09 8.93E-09 3.67E-09 30 Boiler 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00E+00 0.00E+00 32 Main feedwater pumps 5.03E-07 9.24E-07 4.21E-07 0.00E+000.00E+00 0.00E+00 33 Turbine Generator (T/G) Exciter 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00E+00 0.00E+00 34 T/G Hydrogen 8.46E-07 1.78E-06 9.33E-07 0.00E+000.00E+00 0.00E+0035 T/G Oil 3.06E-07 6.15E-07 3.09E-07 0.00E+000.00E+00 0.00E+00 33/34
/35 Catastrophic
 
TG Fire 4.36E-07 9.27E-07 4.91E-07 0.00E+000.00E+00 0.00E+00 5/6/7 Combined CB/Aux/RB Transient
 
Bins 3.11E-06 7.15E-06 4.03E-06 4.78E-07 1.10E-06 6.20E-07 31/36
/37 Combined Turbine Building Transient
 
Bins 1.21E-06 4.64E-06 3.43E-06 0.00E+000.00E+00 0.00E+00  to ULNRC-05876
 
Page 102 of 136 Table 4 - Fire CDF Results by Ignition Frequency Bin Bin # Ignition Source LAR CDF RAI 9b CDF CDF Increase (/yr) LAR CDF (/yr) RAI 9b CDF (/yr) CDF Increase (/yr) 11/24
/25 Combined Plant-Wide Transient Bins 1.03E-06 3.09E-06 2.06E-06 1.68E-08 5.03E-08 3.35E-08 Table 5 - Fire LERF Results by Ignition Frequency Bin Bin No. Ignition Source LAR LERF RAI 9b LERF (/yr) LERF Increase (/yr) LAR LERF (/yr) RAI 9b LERF (/yr) LERF Increase (/yr) 1 Batteries 1.30E-09 4.98E-09 3.68E-09 0.00E+00 0.00E+00 0.00E+00 2 Reactor Coolant Pump 7.52E-10 1.95E-09 1.20E-09 7.58E-12 1.97E-11 1.21E-11 3 Transients
 
and hot work 8.07E-10 2.03E-09 1.22E-09 4.18E-10 1.05E-09 6.34E-10 4 Main Control Board 1.76E-08 5.26E-08 3.51E-08 1.76E-08 5.26E-08 3.51E-08 8 Diesel Generators 1.91E-12 3.43E-12 1.51E-12 0.00E+00 0.00E+00 0.00E+00 9 Air Compressors 3.62E-12 6.63E-12 3.01E-12 0.00E+00 0.00E+00 0.00E+00 10 Battery Chargers 5.51E-09 1.43E-08 8.82E-09 0.00E+00 0.00E+00 0.00E+00 12 Cable Run (Self-ignited cable fires) 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 13 Dryers 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 14 Electric motors 7.92E-11 1.54E-10 7.44E-11 4.45E-11 8.62E-11 4.17E-11 15.1 Electrical
 
Cabinets Non-HEAF 1.62E-07 6.44E-07 4.83E-07 1.03E-08 4.11E-08 3.08E-08 15.2 Electrical
 
Cabinets-HEAF 4.09E-09 1.06E-08 6.52E-09 5.23E-10 1.36E-09 8.35E-10 16.1 Bus Ducts 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 16.2 Iso-phase
 
Bus Ducts 1.25E-09 1.62E-08 1.50E-08 0.00E+00 0.00E+00 0.00E+00    to ULNRC-05876
 
Page 103 of 136 Table 5 - Fire LERF Results by Ignition Frequency Bin Bin No. Ignition Source LAR LERF RAI 9b LERF (/yr) LERF Increase (/yr) LAR LERF (/yr) RAI 9b LERF (/yr) LERF Increase (/yr) 17 Hydrogen Tanks 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 18 Junction box 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 19 Misc.
Hydrogen Fires 4.71E-13 1.22E-12 7.52E-13 4.71E-13 1.22E-12 7.52E-13 20 Off-gas/H2 Recombiner (BWR) 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 21 Pumps 2.34E-08 3.39E-08 1.05E-08 6.80E-11 9.87E-11 3.07E-11 22 RPS MG sets 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 26 Ventilation Subsystems 3.65E-09 6.20E-09 2.55E-09 1.35E-10 2.29E-10 9.43E-11 23 Transformers  6.59E-10 1.06E-09 4.01E-10 3.47E-11 5.59E-11 2.11E-11 27 Transformer-
 
Catastrophic 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 28 Transformer-
 
Non Catastrophic 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 29 Yard transformers (Others) 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 30 Boiler 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 32 Main feedwater pumps 2.50E-09 4.59E-09 2.09E-09 0.00E+00 0.00E+00 0.00E+00 33 Turbine Generator (T/G)
Exciter 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 34 T/G Hydrogen 2.20E-08 4.63E-08 2.43E-08 0.00E+00 0.00E+00 0.00E+00 35 T/G Oil 7.81E-09 1.57E-08 7.89E-09 0.00E+00 0.00E+00 0.00E+00 33/34/35 Catastrophic
 
TG Fire 1.35E-08 2.86E-08 1.51E-08 0.00E+00 0.00E+00 0.00E+00 5/6/7 Combined CB/Aux/RB Transient
 
Bins 8.75E-08 2.01E-07 1.14E-07 9.29E-09 2.13E-08 1.21E-08    to ULNRC-05876
 
Page 104 of 136 Table 5 - Fire LERF Results by Ignition Frequency Bin Bin No. Ignition Source LAR LERF RAI 9b LERF (/yr) LERF Increase (/yr) LAR LERF (/yr) RAI 9b LERF (/yr) LERF Increase (/yr) 31/36/37 Combined Turbine Building Transient Bins 2.19E-08 8.40E-08 6.22E-08 0.00E+00 0.00E+00 0.00E+00 11/24/25 Combined Plant-Wide Transient
 
Bins 2.18E-08 6.53E-08 4.35E-08 2.94E-12 8.81E-12 5.86E-12
 
Some bins may show zero risk contribution in Tables 4 and 5, even though sour ces in those bins exist at Callaway Plant and are included in the fire PRA.
However, if a source only exists in areas that use whole-area burnup and its bin does not have the highest multiplier, then that contribution will not be shown here. For example, there are dryers (bin 13) at Callaway Plant in the Laundry Decontamination Facility (fire area LDF-1), but LDF-1 is a whole-area burnup and it contains some bin 15.1 components. Since bin 15.1 has the highest multiplie r of any applicable bin in LDF-1, the bin 15.1 multiplier is used for the entire LDF-1 ignition frequency, and no contribution from the other, lower-multiplier bins, is shown. The bins that result in the highest increases in fire CDF, CDF, LERF, and LERF are summarized in Table 6. The "theoretical total" sums the increase in each metric with the associated plant-wide total in the current Callaway Plant LAR results.
Table 6 - Bins Causing the Hi ghest Risk Metric Increases Metric Largest Increase Increase LAR Plant-Wide Total Theoretical Total Risk Goal Fire CDF (/yr) 15.1 2.92E-05 2.04E-05 4.96E-05 < 1E-4/yr.
Fire CDF (/yr) 15.1 1.88E-06 1.87E-06 3.75E-06 < 1E-5/yr. Fire LERF (/yr) 15.1 4.83E-07 3.97E-07 8.80E-07 < 1E-5/yr.
Fire LERF (/yr) 4 3.51E-08 3.84E-08 7.35E-08 < 1E-6/yr.
 
Bin 15.1 has three of the four highest risk metric increases, which can be attributed to two primary causes. One, bin 15.1 has many risk-significant ignition sources, so increasing the frequency associated with those sources lead s to a relatively large risk increase. Second, bin 15.1 has one of the highest multiplication factors of al l ignition frequency bins. So, the combination of risk-significant sources and high multiplication factor lead to high contribution to risk increases in this sensitivity study.
 
Bin 4, which is the main control board, contributes notably to delta risk because the "deterministically compliant" risk in the main control room (fire area C-27) is assumed to be zero. As such, any risk increase also causes an equivalent increase in delta risk. Bin 4 also has a relatively high multiplication factor.
to ULNRC-05876
 
Page 105 of 136 As shown in Table 6, even if the bins that contribute the largest risk increases for each risk metric are added to the current plant-wide totals, the new, "theoretical" totals still have considerable margin to the risk goals.
 
The most extreme case in this sensitivity study is assuming that all ignition frequency bins use the 95% confidence frequency simultaneously. Table 7 shows this scenario, which is effectively a sum of all risk metric increases shown in Tables 4 and 5.
Table 7 - Risk Metric Increases with All Bins Using the 95% Frequency Metric Total Increase LAR Plant-Wide Total Theoretical Total Risk Goal Fire CDF (/yr) 4.48E-05 2.04E-05 6.52E-05 < 1E-4/yr.
Fire CDF (/yr) 3.96E-06 1.87E-06 5.83E-06 < 1E-5/yr. Fire LERF (/yr) 8.34E-07 3.97E-07 1.23E-06 < 1E-5/yr.
Fire LERF (/yr) 7.96E-08 3.84E-08 1.18E-07 < 1E-6/yr.
 
As shown, the theoretical totals still show reasonable margin against the risk goals.
 
This sensitivity study increased each ignition frequency bin to its 95% confidence interval value and calculated the plant-wide risk metric increases. The increases were looked at individually (bin-by-bin), as well as a bounding case, in which a ll bins were increased at the same time. Even when all ignition frequency bins are set to their 95% confidence interval ignition frequency, the plant risk metrics preserved adequate margin to the plant-wide risk goals. As such, it can be stated that the Callaway Plant fire PRA has reasonable assurance that statistical variations in the ignition frequency data will not cause significant changes in the risk insights or affect the decision-making processes associated with the transiti on to NFPA-805. 
 
to ULNRC-05876
 
Page 106 of 136 Probabilistic Risk Assessment RAI 10  FAQ 52 (NUREG/CR-6850, Supp. 1) suggests growth times from zero to peak heat release rate (HRR) of 8 min and 2 min, respectively, fo r common trash type fires containe d vs. uncontained within plastic or metal receptacles. These are based on Te sts 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-g al 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 main tenance locale such as the MCR, while Tests 3 and 4, cited by the licensee as the basis for the longer growth time to maximum HRR, were more representative of the type of trash can fire to be expected in at least an occasional maintenance locale. For Tests 7 through 9, the FAQ cites times to initial peak in fire intensity of 7, 8, and 13 min, respectively (i.e., two of the three cited tests support the recommended time of 8 min). Please provide the basis for the assumption of the applicability of Tests 3 and 4, such that the longer 10-min growth time was assumed, including a quantitative estimate of the effect of assuming the appropriate shorter growth time(s).
 
Additional Justification Needed The repeated claim that use of 8 min as a best estimate for the time to reach maximum HRR for a trash can fire is overly conservative should be removed from the response, as it is contradicted by the evidence provided in the RAI, based on FAQ 08-0052.
Note that the shift of the time from 10 to 8 min (only a 20% effect) caused the probabilities of abandoning the MCR to increase from 50% to 135%. Focus the response on the sensitivity evaluation performed, which indicates a very small increase in CDF and should bound the increase in delta-CDF.
Response to Probabilistic Risk Assessment RAI-10
[Note the reference above to NUREG/CR-4860 is meant to refer to NUREG/CR-4680, Heat and Mass Release for Some Transient Fuel Source Fires:
A Test Report (NUREG/CR-4680, SAND86-0312)]
Transient fires in the main control room are more likely to be associated with contained trash can fires than plastic bags (where the latter would be more applicable to Auxiliary Building areas). In this case the applicable FAQ recommended time to maximum HRR is 8 minutes. However, using the data in NUREG/CR-6850 Supplement 1 for the tests that are cited, a best estimate time of 10 minutes was derived and used in the LAR analysis. 
 
In response to this RAI, sensitivity runs were performed on the main control room transient fires to investigate the impact on abandonment time for transient fire reaching peak HRR in 8 min as recommended in FAQ-0052.      to ULNRC-05876
 
Page 107 of 136 A transient fire was postulated for the electrical cabinet area (ECA) and the main control area (MCA). CFAST runs were performed to fi nd the probability of forced evacuat ion prior to suppression for the case with ventilation operable and with ventilation failed. The probability of forced evacuation using an 8 minute and 10 minute time for achieving peak heat release rate are shown in Table 1. Table 2 shows the change in CDF when the 8 minute time is used instead of the 10 minute time.
Table 1 - Difference in Evacuation Proba bilities for 8 and 10 Minute Peak HRR Fire Case (Transient only)
Probability of Forced Evacuation using 10 min to peak HRR Probability of Forced Evacuation using 8 min to peak HRR Transient fire in MCA - ventilation failed 7.79E-4 1.16E-3 Transient fire in MCA - ventilation operable 3.95E-5 9.39E-5 Transient fire in ECA - ventilation failed 0.0 0.0 Transient fire in ECA - ventilation operable 0.0 0.0 This sensitivity study applie s only to transient fires. Transient fires originating in the ECA are shown to be of insufficient strength to cause evacuation. The combination of a high ceiling and limited combustible content lead to a situ ation where it is not possible to create a hot gas layer or cause opacity restrictions. Thus, the sens itivity study has no effect on transient fires occurring in the ECA.
The change in CDF using the 8 min peak HRR as opposed to the 10 min HRR is shown in Table 2.
Table 2 - Increase in CDF Caused by Using 8 min Peak HRR Versus 10 min Peak HRR Case Change in CDF (/yr)Comment MCA-vent operable +7.5E-10 Change is a factor of 135%, but absolute value of CDF is very small. MCA-vent failed 0.0 Transient fire in the MCA will not damage any control panel that affects control room ventilation.
ECA-vent operable 0.0 Due to the volume of the ECA, a transient fire of typical combustible volume will not cause uninhabitable conditions prior to
 
suppression.
ECA-vent failed  0.0 Due to the volume of the ECA, a transient fire of typical combustible volume will not
 
cause uninhabitable conditions prior to
 
suppression.
 
Since transient fires in the ECA cannot lead to evacuation, regardless of the HRR, the change in assumption of peak HRR does not change CDF results. For transient fires in the MCA, only the case    to ULNRC-05876
 
Page 108 of 136 with "ventilation operable" is used. The control cabinets for the CR ventilation are in the ECA and it was assumed the CR ventilation remains operable for transient fires in the MCA. This is a conservative assumption, because the increase in evacuation probability is great est for the "ventilation operable" case. 
 
The change in CDF when using an 8 minutes peak HRR rather than a 10 min peak HRR is +7.5E-10/yr. This increase is insignificant when compared to the entire control room CDF of 7.8E-7/yr. 
 
Probabilistic Risk Assessment RAI 11 Attachment G of the LAR identifies the ASP (RP118B) as a Primary C ontrol Station (PCS). There is then a continuation of a bulleted list which includes numerous indicatio ns 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 cont rols in the list are on RP118B. Otherwise, please explain the apparent discrepancy.
Response to Probabilistic Risk Assessment RAI-11 Response provided by ULNRC-05851 dated April 17, 2012.
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Page 109 of 136 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 N RC staff notes that FSA R 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 pre ssure control response of the pressurizer spray valves and reli ef valves, the status of RCS letdown paths, and assumed operator responses, to justify the difference between the safety analysis of s purious SI and this scenario. In addition, please provide the details of the calculation of the human e rror probability wh ich describe the basis for the time available to perform the action compared to the time to access the manual valve and close it to confirm this action is feasible. The response should justify the assumptions made to bound the time, and the assumptions as to the procedural response to a spuriously open injection flow path (i.e., is a manual actuation of Emergency Core Cooling System (ECCS) required which may further delay the recovery action).
 
Additional Justification Needed Can the MAAP run be conducted using the initial conditions from the FSAR 15.5.1.2 analysis, or the RETRAN run using the initial conditions from the MAAP analysis for S-20?  If so, compare the results to show that the MAAP run is the more accurate representation of the actual scenario. The fact that the FSAR calculation may be more conservative does not necessa rily account for the significant difference in timing (~27 min).
Response to Probabilistic Risk Assessment RAI-12 Description of PRA Scenario The fire scenario in question involv es a fire in the B train class IE switchgear room (fire area C-10) which causes a spurious safety injection signal while at the same time failing the control circuits for the B train Centrifugal Charging Pump (CCP). Both Pressurizer Power Operated Relief Valves (PORV's) are simultaneously failed closed by fire damage. The CCPs will continue to inject, causing pressurizer pressure and level to increase. The pressurizer safety valves may open during the transient to relieve pressure, but initially they will relieve steam and reclose. Eventually, the pressurizer will become water solid and the safety valves will pass water. The safety valve is assumed to fail open once it passes water. This condition becomes an S2 LOCA which requires ECCS injection and recirculation in order to prevent core uncovery. The timing for this sequence was derived from a MAAP run (S-20 in the project MAAP report FAI/10-504 Callaway Plant MAAP 4.0.7 Fire PRA
 
Sequences).
 
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Page 110 of 136 Description of MAAP Analysis The salient features of the MAAP run S-20, from which the timing of this FPRA scenario was derived, are shown in Table 1 below:
 
Table 1 - Parameters for M AAP Run and FPRA Scenario Parameter MAAP Run S-20 C-10 FPRA Scenario  Initiator Spurious Actuation of ECCS at T=0 Spurious SI Signal at T=0 Reactor trip T=0  T=0  Number of Injecting Charging pumps 2, through boron injection header  2, through boron injection header  PORV status Operable Failed  Letdown  Isolated at 5 min  Isolated at T=0  Normal charging pump  Not modeled  Tripped at T=0 Pressurizer heaters Operating  Off  Pressurizer sprays  Not credited  Not credited  Time to PORV Opening  4 min  None  The S-20 MAAP run is designed to show the time the pressurizer would go water solid for scenarios with unmitigated safety injection in the absence of a LOCA event. As seen in Table 1, the MAAP S-
 
20 scenario does not replicate all as pects of the actual FPRA scenario in C-10. In order to be efficient in generating MAAP analysis, the MAAP run S-20 applies to multiple fire scenarios. The MAAP S-20 scenario however, is conservative with respect to the C-10 FPRA scenario, principally because the PORV's are functional in the MAAP run as opposed to failed in the C-10 PRA scenario. This causes a higher reactor pressure throughout the scenario which leads to a lower ECCS flowrate. The MAAP run S-20 assumed the PORV's were operable, so that steam relief from the pressurizer occurred at the set point of the PORV's (2335 psig) ra ther than the set point of the safety valves (2425 psig). The pressurizer pressure would be higher during the actual FPRA scenario (with the failed PORV's),
which would result in a lower charging pump flow rate and a longer time to go water solid, than in MAAP Run S-20. 
 
Explanation of MAAP Results Table 2 explains the critical MAAP features in or der to provide an understa nding of the results.
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Page 111 of 136 Table 2 - MAAP Results for S-20 Parameter Value Total Pressurizer Volume 1854 ft3 Pressurizer Vapor Space during Normal Power Operation 804 ft3 Average ECCS Flowrate (from S-20 results Fig 3-701)  135 gpm Average ECCS Flowrate (from S-20 results Fig 3-702) 126 gpm Normal Letdown Flow 120 gpm Specific volume of water at 652F (pzr water temp)
.02691 Total RWST water outflow, from Fig 3-702 4615 gallons Time to fill pressurizer  36.6 minutes
 
Fig 3-701 shows ESF flowrate in lbm/hr, on a sc ale of 0-83,000. Fig 3-702 shows RWST water level in feet, on a scale of 38.9 feet to 40.3 feet. These two figures are used to calculate how much water is injected into the RCS during the transient. The charging pumps ar e secured at 36.6 minutes in the MAAP run. The flow rates in Table 2 above are interpolations from the figures. Due to uncertainty in reading numbers off a graph, the two figures produce differing values for ECCS flow rate. The outflow of the RWST should be the same as the ECCS flow rate. The volumes and flow rates were converted to gpm assuming properties of water at 100F. The RWST water level curve was deemed more accurate, so the calculation assumes ECCS flow rate was 126 gpm.
 
The free volume of the pressurizer is 804 ft
: 3. The total net inflow into the RCS is 4015 gallons. (This is 4615 gal - minus the 600 gallons that escaped the letdown line until isolated at 5 minutes). The increase in water level occurs in the pressurizer where the temperature is 652F and the specific volume of water is .02691. The additional volume of the 4015 gallons is:
 
  (4015 gal) * (8.3 lb./gal)
* (.02691 specific vol.) = 897 ft
: 3.
The available free volume of the pressurizer is 804 ft 3, which results in a surplus water volume of 93 ft 3. This is interpreted that 93 ft 3 exited the pressurizer PORV when cycling prior to the pressurizer filling with water. 
 
There are configuration differences between the M AAP run S-20 and the fire PRA scenario. These are shown in Table 3, with a qualitative assessment of the effect on the results.
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Page 112 of 136 Table 3 - Scenario Differences between MAAP Run S-20 and FPRA Parameter MAAP Run FPRA scenario Effect on Result Pressurizer
 
heaters  On  Off  The pressurizer heaters will cause the pressure to be higher, thereby reducing charging flow. The MAAP run will produce an optimistic result for the FPRA scenario. PORV  Operable  Failed  The pressurizer is controlled at 2335 psig in MAAP, rather than 2425 psig in the PRA scenario. The MAAP flow rate will be higher, thus providing a shorter time to pressurizer fill. The MAAP run will produce a conservative result for the FPRA. Letdown  Isolated at T=5  Isolate at
 
T=0  Letdown flow is 120 gpm at 120F. By allowing letdown flow to continue for 5 minutes in MAAP , the RCS inventory balance is reduced by  600 gallons. The free surface of the RCS is in the pressurizer which operates at a temperature of 652F, the temperature condition at whic h the 600 gallons must be reconciled. By not isolating letdown until 5 minutes, the
 
MAAP run "artificially" pr ovides an additional 133 ft 3  of pressurizer space. Adjusting the calculation to account for this space would shorten the time to water solid by 133/28.1 = 4.7 minutes.  (28.1 ft 3 is the volumetric flow rate of 120 gpm at pressurizer water conditions).
Comparison with FSAR Analysis 
 
FSAR Chapter 15.5.1.2 describes the plant response to an inadvertent ECCS injection, caused by a spurious SI signal, which shows the pressurizer goes solid in 8.6 mi nutes. The RAI questions why the MAAP result is different from the FSAR case in Chapter 15.5.1.2. 
 
The FSAR analysis is a licensing calculation which uses the RETRAN code and uses conservative initial conditions. The MAAP code is a best estimat e code and uses best estimate initial conditions. A comparison of salient parameters and results between the FSAR run and the MAAP run are shown in Table 4. to ULNRC-05876
 
Page 113 of 136 Table 4 - Comparison of MAAP and RETRAN Parameters Parameter MAAP Run S-20 FSAR 15.5.1.2 Initiator Spurious Actuation of ECCS at T=0 Spurious SI Signal at T=0 Reactor trip T=0 T=0 Number of injecting
 
Charging pumps 2, through boron injection header 2, through boron injection header. Normal charging pump for first 6 minutes PORV status Operational Operational after 9 minutes Letdown Isolated at 5 min N/A Normal charging pump Not in model Operating for 6 minutes Pressurizer heaters Operating Off Pressurizer sprays Not credited Operating Pressurizer water solid 36.6 min 8.75 min Nominal flow rate into RCS 126 gpm 346 gpm for 6 minutes 299 gpm thereafter The differences between FSAR modeling and MAAP modeling cannot be compared on a specific basis without recourse to each code with the ability to do sensitivity studies. The decisive difference is the ECCS flow rate. These dictate pressurizer fill rate. The MAAP run is 126 gpm versus the 346 gpm for RETRAN. It is emphasized that the MAAP run is a best estimate calculation while the FSAR calculation is intenti onally conservative.
Operator Action 
 
The NFPA 805 recovery action human error probability (HEP) in this sequence is calculated to be 1.9E-2. This action was walked down and shown to be feasible and accessible within the time constraints of the scenario. The HRA shows a time delay of 25 minutes until the procedural cue is reached. Execution time is 7 minutes, which includes 2 minutes of travel time. No recovery is credited on either the cognitive or ex ecution portion of the HEP.
 
Sensitivity Study with No Recovery Action
 
It is not feasible to run compar ison studies of this scenario in MAAP and RETRAN. The codes are sufficiently different that the fina l numbers would not be expected to be similar, even for identical initial reactor conditions. Therefore, as a final asp ect of the analysis, a sensitivity study was performed which removed credit for this recovery action in all fire scenarios where it is credited. The results are shown in Table 5. The changes in CDF/LERF/delta CDF/delta LERF are small and do not affect the overall risk profile. 
 
Table 5 - Change in CDF/LERF with No Recovery Action Metric  CDF (/yr) LERF (/yr) Change in CDF/LERF  +1.40E-7  +2.10E-9 Change in CDF/LERF  +1.28E-7  +1.90E-9 to ULNRC-05876
 
Page 114 of 136 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 di scuss 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 scen arios, a description of th e scenario, the locations of the target cables in terms of phys ical separation between the fire s ource and the two targets, and the total risk reduction which would be available if this failure mode were eliminated. In addition, please describe the operator recovery action in terms of its complexity, the time available to complete the action before reaching an unrecoverable condition, and in the context of each fire scenario with regards to other local recovery actions which might be required. A discussion of the risk importance of this recovery action should also be provided in terms of the change in risk if the action were assumed to be unsuccessful.
 
Additional Justification Needed Do the calculational results (Tables 2 through 7) reflect removal of cred it for CPTs in all scenarios?  The discussion following Table 5 suggests that some CPT credit was retained (i.e., use of 0.4 for spurious operation probability). How would the results change if this credit were removed entirely?  Note that, for the delta calculations, the credit should be removed in both the base and comparative cases.
Response to Probabilistic Risk Assessment RAI-13 The RAI has four specific parts as in dicated below. Each aspect is discussed in its own section. Parts 2 and 4 require a sensitivity study.
The sensitivity study was performed with and without credit for CPT's in probability of spurious operation (issue from PRA RAI-9a). Probabilities for spurious closure of BBHV8351A-D were retained at their base case values in this sensitivity study as in PRA RAI-9a. Responses to part 2 and 4 therefore have two sets of sensitivity results.
: 1) The response should address the frequency of fire scenarios, a descripti on of the scenario, the locations of the target cables in terms of physical separation between the fire source and the two targets.
: 2) Total risk reduction which would be available if this failure mode was eliminated. 
: 3) Please describe the operator recovery action in terms of its complexity, the time available to complete the action before reaching an unrecovera ble condition, and in the context of each fire scenario with regards to other local r ecovery actions which might be required. 
: 4) A discussion of the risk importance of this reco very action should also be provided in terms of the change in risk if the action were assumed to be unsuccessful.
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Page 115 of 136 Part 1: Description of Scenario and Where it Occurs The pressurizer has two power operated relief valv es (PORV). The PORV is a DC power operated valve. Each PORV discharges to the Pressurizer Relief Tank through a dedi cated line. Each PORV line is supplied with a motor operated isolation va lve (i.e. block valve), which is powered by 480vac power. The power for the PORV and the block valve are from the same train. Thus, cables for the PORV and the block valve on the same discharge line are routed in the same fire areas. If sufficient cabling exists in a fire area such that a Pressurizer PORV could fail spuriously open and its associated block valve could also fail as-is open, a VFDR was assigned to that fire area in the Nuclear Safety Capability Assessment (NSCA) analysis. Fire Areas with such a VFDR ar e listed in Table 1.
Table 1 - Fire Areas with "PORV-LOCA" VFDRs Fire Area VFDR-ID Recovery Action Credited A-8 A-08-003 No A-11 A-11-001 No A-16 A-16-SOUTH-001 No A-17 A-17-001 No A-18 A-18-003 No A-27 A-27-003 No C-18 C-18-002 No C-21 C-21-003 YES C-22 C-22-003 YES C-23 C-23-002 No C-24 C-24-002 YES C-27 C-27-026 YES C-27 C-27-029 YES C-30 C-30-002 No C-33 C-33-002 No RB-1 RB-03-001 / RB-05-001 YES  The PORVs are designed to fail closed on loss of DC power. The recovery action (RA) assigned to these VFDRs is to locally de-energize the PORV control circuit. A recovery action is not being credited in all fire areas in wh ich the VFDR occurs as further explained below. For the "non-RA areas", the fire risk quantification reflects the risk of not recovering RCS integrity for fires that cause spurious PORV actuation and block valve failure. The risk has been shown to be acceptably low, which can be attributed to one of several reasons:
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Page 116 of 136 a) VFDRs are assigned on a Fire Ar ea wide basis. The use of fire modeling would likely demonstrate that few, or even no, fire scenario s have sufficient cable damage to lead to the VFDR. b) For fire scenarios which have the VFDR, sufficient plant systems to mitigate the PORV-LOCA are available and the risk of core damage is very low.
c) Due to low fire frequency and fixed suppression systems, the initiating event frequency for the scenario is very low.
The fire initiating event frequency for those s cenarios which have the PORV-VFDR are shown in Table 2. Table 2 - Initiating Event Frequency of Fire Scenarios with PORV-VFDR Fire Area Fire Scenario Scenario Initiator Frequency (per year) Scenario Description 1 A-8 (none) 2 n/a n/a A-11 1335-3 9.94E-08 Transient Fi re (suppression failed)
A-16 (none) 2 n/a n/a A-17 A17TS1 1.33E-05 Transient Fire A-17 A17TS4 1.27E-06 Transient Fire A-18 1410-3 9.01E-07 RJ159 (BOP computer)
A-27 (none) 2 n/a n/a C-18 3419-6 5.95E-07 Transient Fire C-18 3419-8 5.95E-07 Transient Fire C-18 3419-9 1.09E-07 Transient Fi re (suppression failed) C-21 3501T15 5.03E-08 Transient Fire C-21 3501T16 7.76E-09 Transient Fire C-21 3501T18 1.92E-07 Transient Fire C-21 3501TXX 8.46E-08 Transient Fire (suppression failed) C-22 3801T2 4.17E-07 Transient Fire C-22 3801T3 5.83E-07 Transient Fire C-22 3801T10 6.38E-08 Transient Fire C-22 3801T14 1.25E-07 Transient Fire C-22 3801TXX 1.25E-07 Transient Fire (suppression failed) C-23 3505-5 1.09E-07 Transient Fi re (suppression failed) C-24 3504-1 1.19E-06 Transient Fire C-24 3504-3 1.19E-06 Transient Fire C-24 3504-5 1.09E-07 Transient Fi re (suppression failed)    to ULNRC-05876
 
Page 117 of 136 Table 2 - Initiating Event Frequency of Fire Scenarios with PORV-VFDR Fire Area Fire Scenario Scenario Initiator Frequency (per year) Scenario Description 1 C-27 RL001/02 3.46E-06 MCB RL1/2 (No Evacuation) C-27 RL021/022 3.46E-06 MCB RL21/22 (No Evacuation) C-27 RL1-2-3-4 3.98E-07 MCB Panel Propagation Scenario (No Evacuation) C-27 RL21-22-23-24 3.98E-07 MCB Panel Propagation Scenario (No Evacuation) C-30 3617-4 1.19E-06 Transient Fire C-30 3617-5 1.09E-07 Transient Fi re (suppression failed) C-33 3804-6 3.44E-07 Transient Fire C-33 3804-7 3.44E-07 Transient Fire C-33 3804-19 1.09E-07 Transient Fi re (suppression failed) RB-1 RB3-T1 2.89E-05 Transient Fire RB-1 RB3-T2 1.88E-05 Transient Fire RB-1 RB5-T6 3.23E-06 Transient Fire RB-1 RB5-T7 4.26E-06 Transient Fire Totals 8.62E-05/yr.
Note 1 - A detailed description of each fire is provided in Attachment 6 of the Detailed Fire Modeling Re port for each Fire Area Note 2 - No fire scenarios were identified that have the VFDR As noted, a detailed description of each fire scenario is provided in Attachment 6 of the detailed fire modeling report for each fire area. Attachment 8 of the same document lists the targets for each fire scenario and the time to damage for each target. The targets listed in Attachment 8 are raceways, and the cables in each raceway can be found in the Nuclear Safety Capability Analysis (Calculation KC-26), as well as the SAFE database.
 
Part 2: Risk Reduction Worth if the PORV-Block Valve Commonality Were Eliminated:
 
Table 3 shows the CDF and delta CDF for each scenario. The first CDF column (called "Scenario CDF with VFDR [RA credited if applicable]") sh ows the CDF of the post-transition plant. This includes core damage arising from the VFDR, with the recovery action (if applicable as shown in Table 1) plus the core damage frequency contribution from all other failures in the scenario. The second CDF column (called "Scenario CDF of comp laint case [no VFDR]")
shows the CDF for the scenario if all VFDR's are eliminated, (i.e. the PORV-VFDR and all others). The third CDF column shows the difference between the two, which is "delta CDF".
 
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Page 118 of 136 For this sensitivity study, the compliant case CDF is used as a bounding estimate of the risk reduction if only the PORV LOCA VFDR failure mode was removed. Since the compliant case removes all
 
VFDRs, it reduces risk to a greater extent th an if only the PORV LOCA VFDRs were removed.
 
If a recovery action is credited for any fire scenario in a fire area, it is credited for all fire scenarios in the fire area.
 
Note that all of the results tables have an "a" and a "b" version due to the combination of this issue with the issue from RAI 9a. The "a" versions show risk results that credit control power transformers (CPT) during the calculation of spurious actuation probabilities based on fire-induced cable damage. The "b" version of each table show risk results with the CPT credit removed. Tables 3a and 3b show the fire CDF reduction if the PORV VFDR failure mode was removed from all scenarios in which it exists.
Table 3a - Bounding CDF Risk Reduction if PORV Failure  Mode was Eliminated (with CPT)
Fire Scenarios CDF with VFDR [RA credited if applicable] (per year) CDF of Complaint Case [no VFDR] (per year) CDF Reduction (per year) All with PORV VFDR 3.37E-06 3.23E-06 1.31E-07
 
Table 3b - Bounding CDF Risk Re duction if PORV Failure  Mode was Eliminated (no CPT)
Fire Scenarios CDF with VFDR [RA credited if applicable] (per year) CDF of Complaint Case [no VFDR] (per year) CDF Reduction (per year) All with PORV VFDR 4.77E-06 4.64E-06 1.36E-07
 
As shown, the total reduction in CDF if the PORV LOCA VFDR were eliminated from the fire areas is approximately 1.31E-7/yr. with CPT credit and 1.36E-7/yr. without CPT credit. Compared to the current plant-wide fire CDF of approximately 2.04E
-5/yr. these reductions are not significant. The current plant-wide fire delta CDF is 1.87E-6/yr. ,which has considerable margin to the Regulatory Guide 1.205 limit of 1E-5/yr. The difference between the "a
" and "b" results for delta risk is very small, because the CPT credit was removed from both the variant condition and the compliant condition. 
 
Tables 4a and 4b shows the LERF equivalent of the results in Tables 3a and 3b.
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Page 119 of 136 Table 4a - Bounding LERF Risk Reduction if PORV Failure  Mode was Eliminated (with CPT)
Fire Scenarios LERF with VFDR  [RA credited if applicable] (per year) LERF of Complaint Case [no VFDR] (per year)
LERF Reduction (per year) All with PORV VFDR 1.91E-07 1.88E-07 3.60E-09 Table 4b - Bounding LERF Risk Reduction if PORV Failure  Mode was Eliminated (no CPT)
Fire Scenarios LERF with VFDR [RA credited if applicable] (per year) LERF of Complaint Case [no VFDR] (per year)
LERF Reduction (per year) All with PORV VFDR 2.47E-07 2.43E-07 3.80E-09
 
As shown in Table 4, the total reduction in LERF if the PORV LOCA VFDR were eliminated from the fire areas is approximately 3.60E-9/yr. with CPT credit and 3.80E-9/yr. without CPT credit. Compared to the current, plant-wide fire LERF of approximately 3.97E-7/yr., th ese reductions are not significant. The current plant-wide fire delta LERF is 3.84E-8/yr., which has considerable margin to the Regulatory Guide 1.205 limit of 1E-6/yr. The difference between the "a" and "b" results for delta risk is very small, because the CPT credit was removed from both the variant condition and the compliant condition.
 
Part 3: Description of Op erator Recovery Action:
 
The operator recovery action for the selected VFDRs involves removal of DC power from the PORV control circuit for BBPCV0455A or BBPCV0456A. This action invol ves opening a 125VDC breaker (NK5108 for BBPCV0455A or NK4421 for BBPCV0456A). NK51 is located in fire area C-16 and NK44 is located in fire area C-15. C-15 and C-16 are the battery rooms for their respective trains.
Removal of DC power will cause th e PORV to revert to its closed position. The recovery action was evaluated in the Task 11 Human Reliability Report (Callaway Fire PRA Report 17671-011). Attachment B of that report provides the details of the calculation. A summary of the salient results is provided below. For all areas in which this action is credited, it was verified by walkdowns that a fire free path exists from the control room to the battery rooms. MAAP analysis was performed to determine the timing required to prevent core uncovery, assuming no ECCS was available. The assumption of no ECCS was used as a bounding case. Although one charging pump is guaranteed available in all sequences, the Bor on Injection Header Isolation valves do not have a function to open in Nuclear Safety Capability Assessment (NSCA) and thus it is possible for some scenarios that they are both failed closed. The critical time for ope rator action is 1.83 hours to prevent core uncovery.
 
The operator action is modeled as two basic events-a cognitive error (OP-COG-FO-PORV) and an execution error (OP-OMA-FF-ISPORV). The HEPs assigned to each action are:
 
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Page 120 of 136 Operator action HEP OP-COG-FO-PORV 9.4E-04 OP-OMA-FF-ISPORV 3.4E-05 TOTAL 9.75E-04
 
The operator actions that are credited in each fire area are considered within the context of all actions that may be required for a fire within that fire area. That is, the feasibility analysis, which is documented in the Nuclear Safety Capability Assessment calculation (KC-26), was performed by considering all actions that are required in each fire area. Then, the most limiting timeline from all of the applicable fire areas was selected as the basis for the final human error probability (HEP) calculation in the 17671-011 analysis.
 
Part 4: Risk Achievement Worth of the Recovery Action:
 
This RAI requested an evaluation of the effect of not crediting the recovery action in any scenario.
Tables 5 and 6 show that the results of not crediting the recovery action in any fire area is an increase in delta CDF of greater than 1.00E-7/yr. This increase was above the Ca llaway Plant NFPA 805 project guidelines for risk increase which led to the crediting of the action in several areas. The increase in fire risk if the local action to close the PORV is always failed was evaluated only for fire areas in which the recovery action is credited. These areas are indicated in Table 1, all other areas shown do not credit the recovery action, so it is already assumed to fail. To calculate the change in risk if the recovery action is always failed, the event OP-OMA-FF-ISPORV was set to fail (event value =
1.0) in the concatenated cutset equation files.
These results are shown in Tables 5a and 5b. 
 
Table 5a - Change in Risk with OP-OMA-FF-ISPORV = 1.0 (with CPT)
Baseline Risk ISPORV Failed Change CDF (per yr.) LERF (per yr.) CDF (per yr.)LERF (per yr.) CDF (per yr.) LERF (per yr.) 3.37E-06 1.95E-07 4.23E-06 2.01E-07 8.68E-07 6.70E-09 Table 5b - Change in Risk with OP-OMA-FF-ISPORV = 1.0 (no CPT)
Baseline Risk ISPORV Failed Change CDF (per yr.) LERF (per yr.) CDF (per yr.) LERF (per yr.) CDF (per yr.) LERF (per yr.) 4.77E-06 2.51E-07 5.67E-06 2.58E-07 9.01E-07 7.60E-09
 
As shown in the tables, removing all credit for the PORV LOCA recovery actions leads to CDF increases of approximately 9E-7/yr. and LERF increases of approximately 7E-9/yr. The increases are slightly larger when credit for CPTs is removed from the spurious actuation probability calculations. 
 
All of the risk increase is delta risk, since the risk increases are due to removing credit for an NFPA 805 recovery action that is mitigating a VFDR. Since delta risk due to VFDRs is directly applicable to
 
the Callaway Plant NFPA 805 application, these risk increases (which are equi valent to delta risk increases) are compared to plant-wi de delta risk in Tables 6a and 6b.
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Page 121 of 136 Table 6a - Risk Increases Compared to Plant-Wide Delta Risk (with CPT) Plant-Wide ISPORV = 1.0 Theoretical Total  CDF (per yr.)  LERF (per yr.)
CDF/CDF Increase (per yr.)
LERF/LERF Increase (per yr.)  CDF (per yr.)  LERF (per yr.) 1.87E-06 3.84E-08 8.68E-07 6.70E-09 2.74E-06 4.51E-08 Table 6b - Risk Increases Compared to Plant-Wide Delta Risk (no CPT) Plant-Wide ISPORV = 1.0 Theoretical Total  CDF (per yr.)  LERF (per yr.)
CDF/CDF Increase (per yr.)
LERF/LERF Increase (per yr.)  CDF (per yr.)  LERF (per yr.) 1.87E-06 3.84E-08 9.01E-07 7.60E-09 2.77E-06 4.60E-08 The differences between the results for the "a" and "b" case for delta risk is very small because the CPT credit was removed from both the variant condition and the compliant c ondition. The total delta risk in Tables 6a and 6b still have reasonable margin to the plant-wide delta risk limits of 1E-5/yr. and 1E-6/yr. for CDF and LERF, respectively.
 
Summary:
The "Pressurizer PORV LOCA scenario" has been shown to occur at a relatively low frequency in Table 2. A bounding assessment of the possible risk reduction if the PORV LOCA failure mode was completely eliminated at the Callaway Plant was s hown in Tables 3a/b and 4a/b. Even if the PORV LOCA VFDR failure mode was completely eliminated from the Callaway Plant, the overall risk reduction would not affect the risk insights or conclusions of the Callaway Plant NFPA 805 Application Request. Most fire areas that have the PORV LOCA VF DR do not credit an NFPA 805 recovery action to reduce risk, as shown in Table 1. Table 5a/b shows the change in risk if that PORV LOCA recovery action credit was removed (i.e. if ISPORV is assumed to fail). Table 6a/b summarizes the change in risk if ISPORV is failed, and adds that risk increase to the plant-wide delta risk totals to create new "theoretical total" delta risk metrics. If ISPORV is assumed to fail in all fire areas, the resultant total plant delta risk is shown to maintain significant margin to the plant-wide delta risk limits of 1E-5/yr. and 1E-6/yr. for CDF and LERF, respectively.
 
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Page 122 of 136 Probabilistic Risk Assessment RAI 14 In the LAR, reference is made to self approval in regard to the NFPA 805 tr ansition results. This is incorrect as self approval thres holds do not apply for the transiti on aspects of the application, but rather are only applicable post-transition in evaluati ng 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 statem ent has any effect on the LAR.
Response to Probabilistic Risk Assessment RAI-14 Response provided by ULNRC-05851 dated April 17, 2012.
 
Probabilistic Risk Assessment RAI 15 Table W-1 of the LAR includes in its title "95% of Calculated Fire CDF," but the table only includes approximately 58 percent of the Fire CDF. Please provide clarification regarding the title and table and revise as appropriate.
 
Response to Probabilistic Risk Assessment RAI 15 Response provided by ULNRC-05851 dated April 17, 2012.
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Page 123 of 136 Probabilistic Risk Assessment RAI 16 Scenarios 3801T3 and 3801T2 (upper cable spreading room transient fires) a ppear 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 ade quate defense-in-depth. A discussion of key assumptions which impact these scenarios conservatively (if any) as well as administrative controls or other measures which re duce the likelihood of transi ent combustibles in the critical location should also be provided. A more accurate quantitative assessment of CCDP should be provided, or a justification as to why this is not possible.
 
Additional Justification Needed The second to last sentence in the 1st paragraph in cludes the sentence, "Duri ng the FRE process, these conservatisms were discussed and the decision was made to retain the conservatisms in the fire PRA and attribute them to "defense-in-depth."  Conservatisms in the model are not generally defense-in-depth attributes. The sentence is not needed as the later discussion does a ddress defense-in-depth.
 
Response to Probabilistic Risk Assessment RAI 16 These scenarios occur in room 3801, which is the Upper Cable Spreading room (Fire Area C-22). Scenarios in the upper cable spreading room are subject to multiple analytical conservatisms that tend to increase the conditional risk calculation results. The total CDF for scenarios 3801T3 and 3801T2 is 7.62E-7/yr. The delta CDF is 9.06E-8/yr. While thes e two scenarios are the highest CDF contributors in Fire Area C-22, the absolute re sult is acceptable with respect to meeting the risk criteria in Regulatory Guides 1.205 and 1.174. There are several conservatisms inherent in these scenarios, which if removed would reduce the fire risk. During the FRE process, these conservatisms were discussed and the decision was made to retain the conservatisms in th e fire PRA and attribute them to "safety margin". These conservatisms are summarized below:
1st Conservatism: Additive Propagation of High Basic Event Probabilities for Seal Failure
 
The dominant cutsets in this scenario involve loss of all RCP seal cooling due to spurious isolation of the seal injection lines and the Component Cooling Water [CCW] thermal barrier return line. The total quantified probability for loss of all RCP seal cooling is 3.1. This is due to the extraordinarily high probabilities postulated for spurious operation of valv es. The top 4 cutsets for loss of seal cooling are 0.32 each. The next 8 cutsets are 0.16 each. The WinNUPRA code adds each cutset without regard to their exclusivity. The WinNUPRA code does not have a quantification capability to provide a "min-cut-upper-bound" result that adjusts for cutsets with high failure probabilities. If the numerical probability of the loss of seal cooling is capped at 1 (guaranteed failure) then the Conditional Core Damage Probability [CCDP] for the entire scenario is reduced to 0.25. to ULNRC-05876
 
Page 124 of 136 2nd Conservatism Additive Propagation of High Basic Event Probabilities for SG Blowdown
 
A similar effect occurs with the steam generator bl owdown isolation valves failing open. Each valve is attributed a 0.8 probability of spurious opening. The AFW (decay heat rem oval) success criteria requires heat removal from two Steam Generators [S G's]. If each SG has a pr obability of failure of 0.8, then the total probability for Loss of AFW is 1.6. In this scenario, success could be achieved by feeding one SG with the A train Auxiliary Feedwater [AFW] motor driven pump and one SG with the Non-safety motor driven auxiliary feedwater pump. This success criteria combination is not modeled in the PRA analysis, but would be available in the event this fire scenario ever occurred.
 
Removing the calculation-related conservatisms, by altering the individual MOV spurious closure probabilities, the CCDP is reduced from ~0.76 to ~
0.25. If the spurious valv e actuation probabilities are changed to a logical 1 (i.e., guaranteed failure), the CCDP redu ces to about 0.35, simply due to elimination of non-minimal cutsets. The point to be made on these scenarios is that there are several valves postulated to spuriously actuate with probabilities which exceed the bounds for "rare-event approximation". The PRA codes show artificially high core damage probabilities when these high probabilities are input. The cutset results were reviewed and considered during the Fire Risk Evaluation [FRE] process and the decision was made to proceed with the over-estimation of CDF. That does not mean the scenario has insufficient defense-in-depth or safety margin.
 
With respect to defense in depth for fire ignition, the upper cable spreading room will be administratively controlled as a "No Transient Combustible Storage" ar ea and also as a "No Hotwork" area while the plant is at power. These controls, plus the absence of any fixed ignition sources, provide reasonable assurance that the likelihood of a challenging fire in the area is very low.
 
In summary, these upper cable spreading room fires have th e following characteristics:
: 1. Low likelihood of a challenging fire, due to no fixed sources and administrative transient controls.
: 2. The area is provided with a wet-pipe suppression system.
: 3. Availability of the "A" Motor Driven Auxiliary Feedwater Pump [MDAFP] and the Non-Safety Auxiliary Feedwater Pump [NSAFP], which are individually failed by spuriously open SG Blowdown isolation valves, but the potenti al for success with both pumps in operation, is not considered. 
: 4. Statistical over calculation of multiple events with high probabilities for spurious operation cause an over estimation of the CCDP. 
 
These characteristics are considered to be sufficient justification that, even with high calculated conditional core damage probabilities; there is reasonable defense in depth available for these fire scenarios and the remaining mitigation capabilities provide an overall risk that is sufficiently low.
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Page 125 of 136 Probabilistic Risk Assessment RAI 17 The change in risk or delta-risk (risk) for fire areas A-30 and TB-1 is identified as "0.00+00" while for fire area C-35 it is identified as "epsilon." Please clarify the intended difference between these table entries. Additional Information Needed The response explicitly defines "epsilon" but not 0.00+00. The difference between the two remains unclear and seems to overlap. The response sh ould explicitly state when 0.00+00 is used
 
Response to Probabilistic Risk Assessment RAI 17 The table below shows the three situa tions for reporting delta risk. The "" (epsilon) is used when the fire ignition sources in the fire area do not cause any damage to Fire PRA components or cables. Although the Fire PRA will not show any risk, a very small risk is possible such as a risk level below the truncation probability of the PRA. As such, these "no fire-induced damage" scenarios are considered to have very small risk and are denoted with the symbol "". SITUATION DELTA RISK IS SHOWN AS (from LAR Table W-2) Fire Area has no VFDRs from NSCA (fire area is deterministically compliant).
N/A Fire Area contains VFDR(s), but detailed fire scenario modeling does not damage any Fire PRA cables or components. Fire Area contains VFDR(s) and fire modeling shows fire damage to PRA cable(s) or component(s). Fire damage may or may not include damage to VFDR cables or components.
Calculated delta risk is shown, which may be zero or non-zero
 
A zero delta risk is only used for two fire areas in Table W-2, A-30 (Auxiliary Feedwater Valve Compartment, SGs "B" & "C") and TB-1 (Turbine Building). These fire areas have "zero" delta risk reported because the VFDR's assigned to these ar eas are not capable of being evaluated under the PRA success criteria. Thus, although NSCA identifies VFDR's, they are quantified in the PRA as having 0.00E+00 risk.
 
A-30 has a VFDR for the Steam Generator-C PORV spuriously opening (VFDR-A-30-001). SG-C is a non-credited steam generator in NSCA, but this is a VFDR because it causes RCS overcooling and interferes with pressurizer water level control. Fire PRA does not model pressurizer water level control, so this has no effect on the PRA.
 
The first two VFDRs for TB (TB-01-001 and TB-01-002), are related to cable damage to various Pressurizer heater power and control power supply cables. The FPRA does not model Pressurizer heaters because they are used to control and maintain sub-cooling in the RCS. The thermal-hydraulic    to ULNRC-05876
 
Page 126 of 136 analyses upon which the PRA success criteria are ba sed show that sub-cooling is not required to mitigate core damage during the postulated accident scenarios in the Internal Events and Fire PRA models. As such, the Pressurizer heaters are not included in the PRA models and the state of the heaters has no impact on calculated risk.
 
VFDR TB-01-003 is a loss of switchgear room "B" HVAC, whic h is not required in the FPRA for success. Even without HVAC for the duration of the FPRA mission time, room temperatures remain low enough to allow unaffected performance of compone nts in the Switchgear Rooms. As such, a loss of Switchgear Room HVAC does not impact the plant's ability to mitigate an accident.
Probabilistic Risk Assessment RAI 18 The disposition of VFDRs with regards to defense-in-depth and safety margins in the LAR in Attachment C provides no technical justifications but simply states an evaluation was performed and found to be acceptable. Please describe the process that was applied to evaluate the acceptability of defense-in-depth and safety margins for VFDRs. (T his should be a general de scription of the process and criteria, not a detailed basis for each VFDR.)
The description should al so address how reliance upon multiple, time-critical, or complex recovery actions for a particular fire sc enario is evaluated to assure there is no over re liance upon operator acti ons, and how the risk evaluations for recovery action probabilities consider multiple actions in a single scenario.
 
Response to Probabilistic Risk Assessment RAI 18 Response provided by ULNRC-05851 dated April 17, 2012.
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Page 127 of 136 Probabilistic Risk Assessment RAI 19 There is no description of how the change-in-risk is estimated for the various VFDRs. Please provide a description about the modeling of the cause-and-effect relationship in the PRA for each type of VFDR (e.g., cable separation issues, degraded barriers). The description should also include any key assumptions or conservatisms in these evaluations including, for example, if recovery actions are included.
 
Additional Information Needed When calculating delta risk between cases (b) and (c), i.e., with vs. without credit for an operator action, where (c) is the "ideal" compliant case (lowest risk), does the "nominal" HEP in case (b) include fire effects (and is therefore hi gher), leading to a greater delta risk?
Response to Probabilistic Risk Assessment RAI 19 Preparing For Delta Risk Calculation VFDR's are identified by the NS CA, based on complete burnout of a FIRE AREA. A VFDR exists in the AREA if any NSCA safety f unction cannot be provided because sufficient trains of a mitigating system are failed by the fire. These VFDR's are then assessed by the PRA for calculation of delta risk.
The first step is for the PRA to determine if the NSCA safety criteria are refl ected in the PRA. Safety functions specified in the NSCA analysis, but not included in th e PRA analysis, have not been assessed for variant risk. These typically involve, for example, sa fety functions associated with maintaining pressurizer water level and RCS subcooling. The PRA does not require maintenance of pressurizer water level or subcooling so these VFDR's do not contri bute to risk because they do not impact functions modeled in the PRA.
 
The next step is to determine if the VFDR compone nts are explicitly modeled in the PRA. An example of components not explicitly modeled in the PRA would be instrumentati on required for operator actions. For these VFDR's, it is nece ssary to identify a surrogate com ponent in the PRA, which can be varied to assess the delta risk. For instruments, the surrogate basic event is generally the human error which requires the instrument as a cue. Alternatively, it could be the component to be actuated by the operator.
 
If a VFDR matches a PRA success criteria and credited system, it is then quantitatively analyzed in the Fire Risk Evaluation process. In these cases, the di sposition of the VFDR is based on a delta risk calculation. The delta risk is calculated as the difference between the variant case (which is the proposed configuration of the "post-transition Callaway plant") compared to the compliant case, which is a hypothetical configuration in which the Callaway plant complies with all deterministic requirements of NFPA-805. This hypothetical, compliant configura tion is referred to as the "deterministic compliant case" in the Callaway Fire Risk Evaluations (FRE) a nd Fire Safety Analysis    to ULNRC-05876
 
Page 128 of 136 (FSA) reports.
 
The majority of VFDRs evaluated in the PRA are cable separation issues. The cable separation VFDRs occur because cables or equipment for both trains of a mitigating system are located in the same Fire Area. Another category of VFDR's involves cable or equipment damage which can put the plant outside of its NSCA mitigation capabilities. These include spurious closure of a RCP seal return isolation valve or spurious opening of a steam generator atmospheric relief valve. These will be
 
VFDR's regardless of where the cables are located.
 
Another type of VFDR is related to degraded fire wrap. In these cases, the risk associated with the VFDR is treated by not giving credit to the degraded wrap in the fire modeling and associated input to the Fire PRA. That is, the cables within the degr aded wrap were treated as though there was no wrap present. Like the cable separation VFDRs, the basi c event(s) corresponding to the VFDR cable(s) are then set to zero to simulate compliance in the deterministically complia nt risk calculation. 
 
A final type of VFDR at the Ca llaway plant is the HDPE piping pr esent in Fire Area C-1, which is potentially vulnerable to fire damage. To treat the risk associated with this VFDR, the HDPE pipe was treated as a potential target in the fire modeling. If the piping was fa iled in a postulated fire scenario, the deterministically compliant risk calculation set the basic events associated with that damage state
 
to zero. Overall, the HDPE piping type of VFDR was treated the same as a cable separation issue.
 
Calculation Of Delta Risk All VFDR's are analyzed for delta risk with the same process. This process is summarized below:
The components associated with the VFDR are identified and the PRA basic events associated with the component failure modes are identified. Recovery actions associated with the VFDR are identified and verified to be in the mode l associated with the variant co mponent. Three risk cases can be calculated, depending on the magnitude of the fire risk:
 
a) The VFDR is modeled as it is in the current plan t configuration with the recovery action set to fail - i.e., this is the current risk with no recovery action.
b) The VFDR modeled with the recovery action at its nominal human error probability, which includes fire effects - i.e., this is the post-transition case.
c) The plant without the VFDR - i.e., assume th e offending cables are separated or wrapped, or otherwise not failing. VFDRs are eliminated from the PRA model by setti ng the basic event(s) that are associated with the offending cable(s) to zero to simulate compliance in the deterministically compliant risk calculation.
 
The difference in risk between case a & c indicates if the VFDR is signi ficant enough to require a recovery action. The difference in risk between case b and case c indicates if the recovery action provides an acceptable risk mitigation strategy. If the delta risk between case b and case c (cumulative for all VFDR's in the plant) does not meet RG 1.205 risk criteria, a plan t change is required.
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Page 129 of 136 Recovery actions for VFDRs are referred to as "805 Recovery Actions" in the Callaway LAR, are not necessarily credited for all VF DRs. If the delta risk between case a & c is small enough, it is acceptable to leave the VFDR in the plant design without a mitigating recovery action.
VFDR's are identified by NSCA on the basis of complete burnout of an entire FIRE AREA. If the fire area is modeled in the PRA as whole room burnout, then the VFDR's are evaluated in a consistent basis with the NSCA. For areas wh ich have fire modeling, the VFDR's are evaluated on the basis of each fire scenario. Whereas an area may have a VFDR in which two cables are located on opposite sides of the room, when fire modeling is employed, th e two cables do not appear in the same scenario, so from the perspective of the PRA, the VFDR does not exist. Or, it is said the VFDR has no quantifiable risk, because it does not appear in any scenario.
 
Probabilistic Risk Assessment RAI 20 Please provide confirmation that the use of the guidance from EPRI TR-1016735, "Fire PRA Methods Enhancements, Additions, Clarifications, and Refinements to EPRI 1011989," included any modifications of this report as incorporated into Supplement 1 of NUREG/CR-6850.
 
Response to Probabilistic Risk Assessment RAI 20 Response provided by ULNRC-05851 dated April 17, 2012.
 
Probabilistic Risk Assessment RAI 21 The disposition for F&O PRM-B4-1 states that the fire-induced risk model report was updated to provide the bases for fire-induced initiators and non-applicability of Supporting Requirement PRM-B4. Please provide these bases.
Response to Probabilistic Risk Assessment RAI 21 Response provided by ULNRC-05851 dated April 17, 2012.
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Page 130 of 136 Probabilistic Risk Assessment RAI 22 The dispositions of F&Os FSS-E03-1 and UNC-A1-2, both related to Supporting Requirement FSS-E3, cite conservatism in method selection and use of data from NUREG/CR-6850 as justification for not meeting the requirement (at Capability Category II) to provide a mean value and statistical representation of uncertainty intervals for parameters used to model significant fire scenarios. Please explain how the requirements of FSS-E3 are met or justify why they need not be.
Additional Information Needed:
The statement that there is no SR to incorporate the fire modeling uncertainties into the CDF/LERF equation uncertainty is misleading, and the conclusion that CC-II is attained questionable (with respect to SR FSS-E3; SR UNC-A1, to which this RAI also a pplies, is either Met or Not Met). As per SR UNC-A1, an uncertainty analysis "in accordance with HLR-QU-E and its SRs in Part 2" must be performed. SR QU-E3 in Part 2 requires an estimation of the uncertainty interval of the CDF results consistent with the characterization of parameter uncertainties for CC-I, with the additional requirement to take into account the state-of-knowledge correlation for CC-II. Thus, it is not sufficient to just discuss the uncertainties on the individual parameters that cont ribute to CDF, but at least an estimate of the uncertainty on CDF itself is required. Provide this estimate. Note also that the term "fire modeling" as used here is not restricted to fire modeling just in the phenomenological sense, such as empirical correlations, zone mode ls, of CDF models, but applies to all the elements that are input into the "fire risk equation" (inc luding ignition frequency, non-suppressi on probability, etc.).  (Note: a roll-up of the various sensitivity analyses performed in response to other PRA RAIs may form a reasonable "estimate" of the CDF uncertainty interval.)
 
Response to Probabilistic Risk Assessment RAI 22
 
Supporting Requirement FSS-E03 of the ASME PRA st andard requires identification of numerical uncertainty bounds for fire modeling parameters in order to meet Capability Category II. Capability Category I allows qualitative assessment of uncertainty. The F&O indicates the Callaway Plant Fire PRA has met Capability Category I, but has not provided numerical uncertainty bounds for fire modeling parameters and thus does not met Capability Category II. 
 
Several sources of uncertainty were consider ed in the fire Modeling. These are discussed quantitatively and qualitatively through the docum ented reports of the Callaway Plant NFPA 805 project. These are discussed in response to Fire Modeling RAI-2a and are equally applicable to this RAI.
 
The discussion and identification of uncertainty bounds is sufficient to attain Capability Category II. There is no SR to incorporate the fire modeling uncertainties into the CDF/LERF equation uncertainty. To provide an estimate of the Fire CDF and LERF, th e various sensitivities that have been performed during the creation of the LAR and the subsequent RAI responses were summarized and compared. The sensitivity studies and their results are shown in the tables below.
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Page 131 of 136 Table 1 - Summary of Fire PRA Sensitivity Studies
# Sensitivity Description 1 FAQ 08-0048 Ignition source bins with  > 1.0 use 6850 mean frequencies instead of FAQ 08-0048 (also discussed in PRA RAI 07f) 2 Thermoplastic Examine the impact of lower damage threshold and self-ignited cable fires for non-IEEE-383 cable in the Turbine Building (also
 
discussed in PRA RAI 07f) 3 PRA RAI 07c Sets certain HEPs with relatively low time window margins to 1.0 4 PRA RAI 08b Removes the reduction factor from bin 16.2 5 PRA RAI 09a Re-calculates plant-wide risk and delta risk without credit for CPTs 6 PRA RAI 09b Uses the 95th% ignition frequencies from NUREG/CR-6850 Supplement 1 for all ignition source bins to study risk sensitivity to ignition frequency 7 PRA RAI 13 Considers the effect of both removing the PORV LOCA VFDR failure mode and assuming that the applicable recovery action
 
always fails
 
Table 2 - Fire PRA Sensitivity Quantitative Results Summary
# CDF Increase (per yr.)
LERF Increase  (per yr.) CDF Increase  (per yr.) LERF Increase  (per yr.)
1 1 3.64E-06 7.77E-08 1.95E-06 4.50E-08 2 8.81E-07 2.29E-08 0.00E+00 0.00E+00 3 3.72E-07 7.91E-09 3.72E-07 7.91E-09 4 8.18E-09 1.59E-10 0.00E+00 0.00E+00 5 6.26E-06 1.48E-07 1.75E-06 7.36E-08 6 4.48E-05 8.34E-07 3.96E-06 7.96E-08 7 7.49E-07 3.53E-09 7.49E-07 3.53E-09 Total Increases 5.31E-05 1.02E-06 6.83E-06 1.65E-07 Plant Totals 2.02E-05 3.97E-07 1.87E-06 3.84E-08 Theoretical Totals 7.33E-05 1.41E-06 8.70E-06 2.03E-07 Risk Goals < 1E-4 < 1E-5 < 1E-5 < 1E-6 Results satisfactory satisfactory satisfactory satisfactory Note 1, sensitivity #1 is considered to be a subset of #6 and, as such, is not included in the sum for the "Total Increase" row.
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Page 132 of 136 Each of these sensitivity studies generally contains conservatism s, as discussed in their corresponding RAIs and/or FPRA report sections. As such, combining all of the risk increases from these independent sensitivity studies cascades the conservatisms to provi de a truly conservative, bounding estimate of fire risk uncertainty. 
 
As noted in Table 2, even if all of the sensitivity study risk increases are summed and compared to the risk goals from Regulatory Guide 1.205, the totals ar e below the regulatory lim its. As described above, these total risk metrics are subj ect to considerable conservatism and can be considered bounding for fire risk uncertainty.
 
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Page 133 of 136 Probabilistic Risk Assessment RAI 23 In some of the Fire Evaluation of Delta Risk Calcul ations for the various fire areas, a HRR profile for a transient combustible less than the recommended (142 kW and 317 kW at the 75th and 98th percentiles) was assumed. Please provide the bases for these assumptions.
 
Additional Information Needed When citing as a basis for limiting the transient combustible HRR to 69 kW (98th %ile) that 69 kW is judged to be no larger than the 75th
%ile fire in an electrical cabinet with qualified cable, further note, if correct, that this assumes no more than one cable bundle involve d in the fire. Table E-1 of NUREG/CR-6850 reports two 75th %ile values for qua lified cable, one at 69 kW (single bundle), the other at 211 kW (multiple bundles).
 
In the response to PRA RAI 23, the following needs a dditional clarification. Th e last bullet in the list of the basis for assuming a peak transient HRR of 69 kW cites the test results from Table G-7 of NUREG/CR-6850, stating that
"... the types of fires that can be expected in these rooms (i.e., polyethylene trash can or bucket cont aining rags and paper) were measured at peak HRRs of 34 kW or below."  This is true except for one test, SNL-No wlen, Test #9, where a 50 kW HRR during the first 15 min was observed. Since this test involved a much larger trash can (30 gal) than any of the others (maximum of 5 gal), the bullet s hould add either of the following cl arifications: (1) "i.e., a maximum-sized polyethylene trash can or bucke t of ~5 gal. containing rags or paper;" or (2) "were measured at peak HRRs of 50 kW or less."  Provide the appropriate clarification.
Response to Probabilistic Risk Assessment RAI 23 EPRI-led Fire PRA Methods Review Panel issued decisions on methods submitted for their review. Letter from NEI to NRC, B. Bradley to D.
Harrison, "Recent Fire PRA Methods Review Panel Decisions: Clarifications for Transient Fires and Alignment for Pump Oil Fires" dated September 27, 2011 provided as a clarification to the guidan ce of NUREG/CR-6850 and part of the new PRA methods. Attachment 1 "Description of Treatment for Transient Fires," and Attachment 3 "Panel Decision," allow the user to choose a lower screening heat release rate for transient fires in a fire compartment based on "the specific attributes and considerations a pplicable to that location." The guidance indicates that "plant admi nistrative controls should be considered in the appropriate HRR for a postulated transient fire" and that "a lower screening HRR can be used for individual plant specific locations if the 317kW value is judged to be unrealistic given the specific attri butes and considerations applicable to that location."
 
At Callaway Plant, a 69kW transient heat release ra te was justified for certain fire areas based on several factors:
 
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Page 134 of 136
* All fire areas which were credited for a reduced heat release rate are subject to strict combustible controls (areas de signated as "No Storage") and so paper, cardboard, scrap wood, rags and other trash shall not be allowed to accumulate in the area.
* Large combustible liquid fires are not expected in these areas since activities in the areas do not include maintenance of oil containing equipment.
* The transient fire history in the plant was revi ewed and a transient fire has not occurred in these fire areas.
* A transient fire in an area of strict combustible controls, where only small amounts of contained trash are considered possi ble, is judged to be no larger than the 75th percentile fire in an electrical cabinet with one bundle of qualified cable.
* The materials composing the fuel packages included in Table G-7 of NUREG/CR-6850 (e.g., eucalyptus duff, one quart of acetone, 5.9kg of methyl alcohol, etc.) are not representative of the typical materials expected to be located in these areas.
* A review of the transient ignition source tests in Table G-7 of NUREG/
CR-6850 indicates that of the type of transient fires that can be expected in these room s (i.e., polyethylene trash can or bucket containing rags and paper) were measured at peak heat release rates of 50kW or below.
 
Since only small quantities of trash in temporary containers can be expected, a 69kW peak heat release rate was determined to be appropriate to represent this quantity of combustibles. The 69kW heat release rate bounds the small trash can fire s reported in NUREG/CR-6850 Appendix G.
to ULNRC-05876 Page 135 of 136 Section 6:  Licensee Identified Changes (LIC) LIC-01 Provided by ULNRC-05851 dated April 17, 2012 LIC-02 Provided by ULNRC-05851 dated April 17, 2012 LIC-03 Provided by ULNRC-05851 dated April 17, 2012 LIC-04 Provided by ULNRC-05851 dated April 17, 2012 LIC-05 Provided by ULNRC-05851 dated April 17, 2012 LIC-06 Provided by ULNRC-05851 dated April 17, 2012 LIC-07 Provided by ULNRC-05851 dated April 17, 2012 LIC-08 Provided by ULNRC-05851 dated April 17, 2012 LIC-09 LAR Table 4-3 and Attachment C have been revised to report Fire Area TB-1 Room 3307 fire suppression system as "Y*" under the "Required for Chapter 4 Separation Criteria" column. The note in Table 4-3 correctly indicates the system is required for Chapter 3 compliance. LIC-10 LAR Attachment J was revised to include positive statements that the models are used within their validated range. Section 4.5.1.2 of the LAR has been updated to provide additional discussion related to the statements made in Attachment J. LIC-11 LAR Table 4-3 and Attachment C have been revised to indicate that fire detection in Fire Zones 1206 and 1207 and Detection Zone 100 in Fire Zone 1130 are required to
 
support NFPA 805 Chapter 4 Separation requirements. LIC-12 LAR Table 4-3 and Attachment C have been revised to remove duplicate fire zones reported in Fire Area TB-1, Fire Zone s 3706 and 3612. The entry for Fire Zone 3706 designated as "Fire Brigade Storage Area" has been removed. The entry for Fire Zone 3612 designated as "Field Office" has been removed. LIC-13 It was not clear to the staff what the deficiency in the fire PRA model is relative to F&O PRM-B2-1 nor if the item was resolved by making changes or by simply
 
clarifying the underlying issue. Changes have been made in Attachment V to address this. LIC-14 Attachment V has been revised to cl arify the basis for closing F&O PRM-B4-1. LIC-15 The NRC has identified that the use of the Electric Power Research Institute (EPRI)
Technical Report TR-1006756 constitutes use of performance based methods for NFPA 805 Chapter 3 requirements. Therefore, a new NRC Approval Request 3 for use of the EPRI document has been added to Attachment L. LIC-16 The LAR Table B-1 Section 3.4.1(a) is revised and a new LAR/TR Attachment T, Clarification of Prior NRC Approval Request 6, has been added to address the 2 hr grace period for Fire Brigade staffing allowed during Operations Main Control Room shift change. The allowance for the 2 hr gr ace period was originally included in the Westinghouse Standardized Technical Specifi cations that were a pproved for Callaway Plant. LIC-17 The LAR Table B-2 Sections 3.5.1.5 and 3.5.2.3 are revised to provide more detail regarding the Callaway Plant method for ev aluation of spurious actuations for ungrounded DC circuits. to ULNRC-05876
 
Page 136 of 136 LIC-18 Ameren Missouri has revised LAR Table B-3 and Table 4-3 for fire areas A-1, A-16, A-27, and C-1 to identify that a 20 ft. separation zone is credited as allowed by NFPA 805 Section 4.2.2.3(b) within the fire area to meet deterministic requirements. This information is added to the Fire Area Comments section of the B-3 table and to the features section of Table 4-3. Additionally, LAR Attachment X, Other Requests for Approval, has been revised to add a ne w request for NRC ap proval of the 20 ft. separation zones in fire areas A-1, A-16, A-27 and C-1. Additionally, the Fire Safety Analysis for each fire area has been updated to enhance the descriptions of the 20 ft.
 
separation zones to be consistent with the level of detail found in Attachment X. Also, LAR Table B-3 and Table 4-3 for fire area RB-1 have been updated to credit the 20 ft. separation zones previously included in LAR Attachment X.      LIC-19 NFPA 805, Section 2.7.3 contains the requirements for Quality Assurance and in response; LAR / TR Section 4.7.3 provided a description of the Post Transition QA Program. This LIC revises that description of the Post Transition QA Program specifically discussing the FP QA Program requirement to conduct independent audits of the FP Program by the Nuclear Oversight Department. The revised text results in
 
two changes: 1) the location of the description of the audits which includes the scope is being removed from the Operating Qualit y Assurance Manual (OQAM) and placed in FSAR Standard Plant (SP) Section 9.5.1 with the FP Program QA requirements such that all FP QA Program requirements will be consolidated within that section of the
 
FSAR, and 2) the frequency at which the audi ts must be conducted as stated in OQAM Section 18 is being revised from 2 years to 3 years. There is no change to the scope of the 3 year audit; it remains the same as stated in the current QA program. Additionally, there is no change to the requirements for personnel conducting the audits. 
 
Attachment A:  Revisions to Transition Report Attachment A - NEI 04-02
 
Table B Transition of
 
Fundamental Fire Protection
 
Program and Design Elements
 
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 TransitionAmeren MissouriCallaway Plant NFPA 805 Transition Repor tNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentSee implementation items identified below.Procedure APA-ZZ-00703, "Fire Protection Operability Criteria and Surveillance Requirements," Rev. 20 /
AllCalculation KC-162, "Performance Based Fire Protection Surveillance Frequency Program," Rev. 0Procedure APA-ZZ-00700, "Fire Protection Program," Rev. 18CAR 201101832, "Track Implementation Items for NFPA-805 Project" / AllComplies, with Required Action3.2.3(1) Inspection, testing, and maintenance for fire protection systems and features credited by the fire protection programIMPLEMENTATION ITEMS:Procedures APA-ZZ-00700 and APA-ZZ-00703 will be revised to include inspection, testing, and maintenance requirements for all f ireprotection systems and features credited by the fire protection program.11-805-048During the implementation of the NFPA 805 license basis, performance-based surveillance frequencies will be established as described in Electric Power Research Institute (EPRI) Technical Report TR-1006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features" and evaluated in Callaway Plant Calculation KC-162, "Performance Based Fire Protection Surveillance Frequency Program."11-805-069NRC approval of the use of EPRI Technical Report TR-1006756 in establishing performance-based inspection, testing, and maintenance frequencies for fire protection systems and features credited by the fire protection program is being requested in Attachment L.NoneSubmit for NRC ApprovalInspection, testing, and maintenance for fire protection systems and features credited by the fire protection programAugust 2011Page A-5 LIC 15 Attachment A. NEI 04-02 Table B Tr ansition of Fundamental FP Program an d Design Elements (NFPA 805 Chapter 3)
Table B NFPA 805 Ch. 3 TransitionAmeren Missouri Callaway Plant NFPA 805 Transition ReportNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentBulk hydrogen complies with the requirements of NFPA 50A-1973 Edition. Exceptions requiring further action are identified below.Calculation KC-27, "NFPA Code Conformance Review," Rev. 0 / Appendix A, Section 50ANFPA 50A, "Standard for Gaseous Hydrogen Systems at Consumer Sites," 1973 Edition / AllCAR 201101832, "Track Implementation Items for NFPA 805-Project" / AllComplies, with Required Action 3.3.7.1 Storage of flammable gas shall be located outdoors, or in separate detached buildings, so that a fire or explosion will not adversely impact systems, equipment, or components important to nuclear safety. NFPA 50A, Standard for Gaseous Hydrogen Systems at Consumer Sites, shall be followed for hydrogen storage.IMPLEMENTATION ITEMS:Procedures will be revised to ensure that the hydrogen supply system is inspected annually and maintained by Ameren Missouri.07-050A-001Dry 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.07-050A-002No Additional ClarificationFSAR Site Addendum (SA), Rev. OL-15 / Section 2.2.2.1.2.1Complies 3.3.7.2 Outdoor high-pressure flammable gas storage containers shall be located so that the long axis is not pointed at buildings.No Additional ClarificationSafe Work Practices Manual, Rev. 18 / "Compressed Gases" SectionComplies 3.3.7.3 Flammable gas storage cylinders not required for normal operation shall be isolated from the system.
August 2011 Page A-25 FPE RAI 01 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 TransitionAmeren MissouriCallaway Plant NFPA 805 Transition Repor tNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentPer Section 6.2.2.e of NUREG-1058, "Technical Specifications Callaway Plant, Unit No. 1," "A site Fire Brigade of at least five members (may be less than the minimum requirements for a period of time not to exceed 2 hours in order to accommodate unexpected absence provided immediate action is taken to fill the required positions) shall be maintained onsite at all times."Per Section 9.5.1.6 of NUREG-0830, entitled "Administrative Controls, Fire Brigade, Technical Specifications, and Training," "The applicant has committed to follow the staff Standard Technical Specifications. The staff finds this acceptable."NUREG-0830, "Safety Evaluation Report Related to the Operation of Callaway Plant, Unit No. 1," dated October 1981 / Section 9.5.1.6NUREG-1058, "Technical Specifications Callaway Plant, Unit No. 1" / Section 6.2.2.eComplies by Previous NRC Approval3.4.1(a) On-Site Fire-Fighting Capability.A fully staffed, trained, and equipped fire-fighting force shall be available at all times to control and extinguish all fires on site. This force shall have a minimum complement of five persons on duty and shall conform with the following NFPA standards as applicable:August 2011Page A-34 LIC 16 Attachment A. NEI 04-02 Table B Tr ansition of Fundamental FP Program a nd Design Elements (NFPA 805 Chapter 3)
Table B NFPA 805 Ch. 3 TransitionAmeren Missouri Callaway Plant NFPA 805 Transition ReportNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentThe industrial fire brigade complies with NFPA 600-2000 Edition. Exceptions requiring further action are identified below.Calculation KC-27, "NFPA Code Conformance Review," Rev. 0 / Appendix A, Section 600NFPA 600, "Standard on Industrial Fire Brigades," 2000 Edition / AllCAR 201101832, "Track Implementation Items for NFPA-805 Project" / AllComplies, with Required Action3.4.1(a)(1) NFPA 600, Standard on Industrial Fire Brigades (interior structural fire fighting)IMPLEMENTATION ITEMS: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 2-2.4.07-600-001Fire 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 authority07-600-002A 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.07-600-003The Callaway Plant Fire Brigade training program will be updated to include a periodic review of NFPA 600.07-600-004A 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-00700.07-805-015 August 2011 Page A-34 FPE RAI 02 FPE RAI 02 Attachment A. NEI 04-02 Table B Tr ansition of Fundamental FP Program a nd Design Elements (NFPA 805 Chapter 3)
Table B NFPA 805 Ch. 3 TransitionAmeren Missouri Callaway Plant NFPA 805 Transition ReportNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentNo Additional ClarificationProcedure APA-ZZ-00912, "Callaway Plant Medical Physical Program," Rev. 16 / Section 4.3Procedure APA-ZZ-01000, "Callaway Radiation Protection Program" (CTSN 4111), Rev. 33 / Section 4.18Complies3.4.1(e) Each industrial fire brigade member shall pass an annual physical examination to determine that he or she can perform the strenuous activity required during manual firefighting operations. The physical examination shall determine the ability of each member to use respiratory protection equipment.See implementation item identified below.Callaway Plant Fire Preplan Manual, Rev. 34 / AllCAR 201101832, "Track Implementation Items for NFPA-805 Project" / AllComplies, with Required Action3.4.2 Pre-Fire Plans.Current and detailed pre-fire plans shall be available to the industrial fire brigade for all areas in which a fire could jeopardize the ability to meet the performance criteria described in Section 1.5.IMPLEMENTATION ITEMS: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.11-805-076 August 2011 Page A-37 FPE RAI 03 Attachment A. NEI 04-02 Table B Tr ansition of Fundamental FP Program a nd Design Elements (NFPA 805 Chapter 3)
Table B NFPA 805 Ch. 3 TransitionAmeren Missouri Callaway Plant NFPA 805 Transition ReportNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentNo Additional ClarificationProcedure APA-ZZ-00700, "Fire Protection Program," Rev. 18 / Section 3.4.8Complies 3.4.2.2 Pre-fire plans shall be reviewed and updated as necessary.See implementation item identified below.Procedure APA-ZZ-00700, "Fire Protection Program," Rev. 18 / Section 3.4.8CAR 201101832, "Track Implementation Items for NFPA-805 Project" / AllComplies, with Required Action 3.4.2.3 Pre-fire plans shall be available in the control room and made available to the plant industrial fire brigade.IMPLEMENTATION ITEMS
:A statement will be added to procedure APA-ZZ-00700 to require that controlled copies of the pre-fire plans be maintained in th e Control Room and made available to the fire brigade.07-805-047The pre-fire plans do not address coordination with other plant groups, this information is contained within the referenced procedures, which are used in conjunction with the pre-fire plans as part of the overall fire response.Procedure OTO-KC-00001, "Fire Response," Rev. 8 / Step 15Procedure EIP-ZZ-00226, "Fire Response Procedure for Callaway Plant," Rev. 14 / Section 5.2Complies with Clarification 3.4.2.4 Pre-fire plans shall address coordination with other plant groups during fire emergencies.N/A - General statement; No technical requirements N/A N/A3.4.3 Training and Drills.Industrial fire brigade members and other plant personnel who would respond to a fire in conjunction with the brigade shall be provided with training commensurate with their emergency responsibilities.
August 2011 Page A-39 FPE RAI 04 FPE RAI 04 Attachment A. NEI 04-02 Table B Tr ansition of Fundamental FP Program a nd Design Elements (NFPA 805 Chapter 3)
Table B NFPA 805 Ch. 3 TransitionAmeren Missouri Callaway Plant NFPA 805 Transition ReportNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentEquipment 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. See implementation item identified below.Procedure APA-ZZ-00700, "Fire Protection Program," Rev. 18 / AllProcedure APA-ZZ-00743, "Fire Team Organization and Duties," Rev. 23 / Section 4.1.3.eProcedure HTP-ZZ-05006, "Fire Involving Radioactive Material or Entry into the Radiologically Controlled Area," Rev. 9 / Section 6.1.2HDP-ZZ-08000, "Respiratory Protection Program," Rev. 21 / Section 3.9.2Calculation KC-27, "NFPA Code Conformance Review," Rev. 0 / Appendix A, Section 600CAR 200902315, "NFPA 805 Transition - Site Organizations Support Tracking CAR" / AllCAR 201101832, "Track Implementation Items for NFPA-805 Project" / AllProcedure APA-ZZ-00700, "Fire Protection Program," Rev. 18 / AllComplies, with Required Action3.4.4 Fire-Fighting equipment.Protective clothing, respiratory protective equipment, radiation monitoring equipment, personal dosimeters, and fire suppression equipment such as hoses, nozzles, fire extinguishers, and other needed equipment shall be provided for the industrial fire brigade. This equipment shall conform with the applicable NFPA standards.IMPLEMENTATION ITEMS: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-00700.07-805-015 August 2011 Page A-44 FPE RAI 05 Attachment A. NEI 04-02 Table B Tr ansition of Fundamental FP Program a nd Design Elements (NFPA 805 Chapter 3)
Table B NFPA 805 Ch. 3 TransitionAmeren Missouri Callaway Plant NFPA 805 Transition ReportNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentN/A - General statement; No technical requirements N/A N/A 3.5 N/AN/A - General statement; No technical requirements N/A N/A 3.5.1 A fire protection water supply of adequate reliability, quantity, and duration shall be provided by one of the two following methods.Callaway Plant complies with subsection (b) to this requirement; therefore, compliance with subsection (a) is not required.
N/A N/A3.5.1(a) Provide a fire protection water supply of not less than two separate 300,000-gal (1,135,500-L) supplies.
August 2011 Page A-46 FPE RAI 11WaterSupply Attachment A. NEI 04-02 Table B Tr ansition of Fundamental FP Program a nd Design Elements (NFPA 805 Chapter 3)
Table B NFPA 805 Ch. 3 TransitionAmeren Missouri Callaway Plant NFPA 805 Transition ReportNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentPer the references identified, the largest design demand of any credited sprinkler or fixed water spray system in the power block is SKC29 at 2300 gpm. The fire flow rate is 2300 gpm + 500 gpm hose stream allowance = 2800 gpm. The total amount of water flowed over two hours would be 2800 gpm x 120 min = 336,000 gallons. Per the references identified, an adequate reliability, quantity, and duration is available to meet this demand.Calculation M-650-00071, "Hydraulic Calculations for Turbine Building EL 2000-0 South, Standardized Nuclear Unit Power Plant System - SNUPPS 10466-M-650," Rev. 1 / AllCalculation 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," Rev. 1C / AllCalculation M-KC-413, "Fire Protection Determines the Flow Requirements of the Fire Pump," Rev. 0 / AllDrawing F/P 095067, "Fire Protection System Fire Water Storage Tank General Plan," Rev. 4 / AllProcedure APA-ZZ-00703, "Fire Protection Operability Criteria and Surveillance Requirements," Rev. 20 / Section 4.3.3.a.1Complies with Clarification3.5.1(b) Calculate the fire flow rate for 2 hours. This fire flow rate shall be based on 500 gpm (1892.5 L/min) for manual hose streams plus the largest design demand of any sprinkler or fixed water spray system(s) in the power block as determined in accordance with NFPA 13, Standard for the Installation of Sprinkler Systems, or NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection. The fire water supply shall be capable of delivering this design demand with the hydraulically least demanding portion of fire main loop out of service.
August 2011 Page A-47 FPE RAI 11 Attachment A. NEI 04-02 Table B Tr ansition of Fundamental FP Program a nd Design Elements (NFPA 805 Chapter 3)
Table B NFPA 805 Ch. 3 TransitionAmeren Missouri Callaway Plant NFPA 805 Transition ReportNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentNo Additional ClarificationCalculation KC-27, "NFPA Code Conformance Review," Rev. 0 / Appendix A, Section 22Drawing F/P 095067, "Fire Protection System Fire Water Storage Tank General Plan," Rev. 4 / AllNFPA 22, "Standard for Water Tanks for Private Fire Protection," 1974 Edition / AllComplies 3.5.2 The tanks shall be interconnected such that fire pumps can take suction from either or both. A failure in one tank or its piping shall not allow both tanks to drain. The tanks shall be designed in accordance with NFPA 22, Standard for Water Tanks for Private Fire Protection.Exception No. 1: Water storage tanks shall not be required when fire pumps are able to take suction from a large body of 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 LIC-01 Attachment A. NEI 04-02 Table B Tr ansition of Fundamental FP Program a nd Design Elements (NFPA 805 Chapter 3)
Table B NFPA 805 Ch. 3 TransitionAmeren Missouri Callaway Plant NFPA 805 Transition ReportNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentNo Additional ClarificationCalculation M-KC-316, "Fire Protection System Hydraulic Calculations Determine th Adequacy of the Fire Protection System for Providing the Design Flow and Pressure to the Interface with the Sprinkler System," Rev. 1C / AllCalculation M-KC-413, "Fire Protection Determines the Flow Requirements of the Fire Pump," Rev. 0 / AllCalculation KC-27, "NFPA Code Conformance Review," Rev. 0 / Appendix A, Section 20NFPA 20, "Standard for the Installation of Centrifugal Fire Pumps," 1974 Edition / AllComplies 3.5.3 Fire pumps, designed and installed in accordance with NFPA 20, Standard for the Installation of Stationary Pumps for Fire Protection, shall be provided to ensure that 100 percent of the required flow rate and pressure are available assuming failure of the largest pump or pump power source.
August 2011 Page A-49 LIC-02 LIC-03 and LIC-05 Attachment A. NEI 04-02 Table B Tr ansition of Fundamental FP Program a nd Design Elements (NFPA 805 Chapter 3)
Table B NFPA 805 Ch. 3 TransitionAmeren Missouri Callaway Plant NFPA 805 Transition ReportNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentPer Section 9.5.1.1 of NUREG-0830, "The water supply system consists of three fire pumps separately connected to a buried, 14-in pipe loop around the plant. There are three 50-percent capacity fire pumps, each rated at 1500 gpm at 347-ft head. One pump is electric motor driven and two are diesel engine driven.""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.NUREG-0830, "Safety Evaluation Report Related to the Operation of Callaway Plant, Unit No. 1," dated October 1981 / Section 9.5.1.1Letter ULNRC-00189 from Bryan (UE) to Rusche (NRC) dated April 15, 1977  / Section 9.5.1.1Complies by Previous NRC Approval 3.5.4 At least one diesel engine-driven fire pump or two more seismic Category I Class 1E electric motor-driven fire pumps connected to redundant Class 1E emergency power buses capable of providing 100 percent of the required flow rate and pressure shall be provided.No Additional ClarificationDrawing 8600-X-88446, "Building Architectural Plan Fire Pumphouse Fire Protection System," Rev. 3 / AllComplies 3.5.5 Each pump and its driver and controls shall be separated from the remaining fire pumps and from the rest of the plant by rated fire barriers.
August 2011 Page A-50 LIC-04 Attachment A. NEI 04-02 Table B Tr ansition of Fundamental FP Program a nd Design Elements (NFPA 805 Chapter 3)
Table B NFPA 805 Ch. 3 TransitionAmeren Missouri Callaway Plant NFPA 805 Transition ReportNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentThe exception to this requirement is utilized at Callaway Plant by providing equipment on two mobile units. Each mobile unit has equipment equivalent to that of three hose houses.Drawing 8600-X-88448, "Fire Loop and Laterals," Rev. 24 / AllCA2112, "Fire Brigade Equipment Inventory and Condition Checklist," dated 1/6/06 / AllCalculation KC-27, "NFPA Code Conformance Review," Rev. 0 / Appendix A, Section 24NFPA 24, "Standard for Outside Protection," 1973 Edition / AllComplies with Clarification3.5.15 Hydrants shall be installed approximately every 250 ft (76 m) apart on the yard main system. A hose house equipped with hose and combination nozzle and other auxiliary equipment specified in NFPA 24, Standard for the Installation of Private Fire Service Mains and Their Appurtenances, shall be provided at intervals of not more than 1000 ft (305 m) along the yard main system.Exception: Mobile means of providing hose and associated equipment, such as 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 FPE RAI 06 Attachment A. NEI 04-02 Table B Tr ansition of Fundamental FP Program a nd Design Elements (NFPA 805 Chapter 3)
Table B NFPA 805 Ch. 3 TransitionAmeren Missouri Callaway Plant NFPA 805 Transition ReportNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentNRC approval is being requested in Attachment L for the use of the fire protection water supply system for purposes other than fire protection.NoneSubmit for NRC Approval3.5.16 The fire protection 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.N/A - General statement; No technical requirements N/A N/A 3.6 N/AStandpipe and hose systems in power block buildings comply with NFPA 14-1976 Edition except as identified below.Calculation KC-27, "NFPA Code Conformance Review," Rev. 0 / Appendix A, Section 14NFPA 14, "Standard for the Installation of Standpipe and Hose Systems," 1976 Edition / AllComplies with Clarification 3.6.1 For all power block buildings, Class III standpipe and hose systems shall be installed in accordance with NFPA 14, Standard for the Installation of Standpipe, Private Hydrant, and Hose Systems.
August 2011 Page A-57 LIC-06StandpipeandHoseStations Attachment A. NEI 04-02 Table B Tr ansition of Fundamental FP Program a nd Design Elements (NFPA 805 Chapter 3)
Table B NFPA 805 Ch. 3 TransitionAmeren Missouri Callaway Plant NFPA 805 Transition ReportNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentPer Section 7-2.3 of NFPA 14-1976 Edition, the valves in the main connection to automatic sources of water supply shall be open at all times. There are motor-operated valves that isolate the containment standpipes, which must be opened manually from the control room to allow water into the containment standpipe risers. Per Page 9.5B-225 of the attachment to SLNRC 81-050, "To protect the chloride sensitive piping and equipment from fire protection system leakage, the standpipes inside the reactor building are normally dry. Control room operator action is required to charge the standpipes. The 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.Calculation KC-27, "NFPA Code Conformance Review," Rev. 0 / Appendix A, Section 14NFPA 14, "Standard for the Installation of Standpipe and Hose Systems," 1976 Edition / Sections 7-2.3 and 7-2.4Letter SLNRC 81-050 from Petrick (SNUPPS) to Denton (NRC) dated June 29, 1981 / Attachment, Pages 9.5B-225 and 9.5E-1NUREG-0830, "Safety Evaluation Report Related to the Operation of Callaway Plant, Unit No. 1," dated October 1981 / Section 9.5.1.6Complies by Previous NRC Approval 3.6.1 For all power block buildings, Class III standpipe and hose systems shall be installed in accordance with NFPA 14, Standard for the Installation of Standpipe, Private Hydrant, and Hose Systems.
August 2011 Page A-58 LIC-05 Attachment A. NEI 04-02 Table B Tr ansition of Fundamental FP Program a nd Design Elements (NFPA 805 Chapter 3)
Table B NFPA 805 Ch. 3 TransitionAmeren Missouri Callaway Plant NFPA 805 Transition ReportNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference 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."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.
August 2011 Page A-59 LIC-05 LIC-06 Attachment A. NEI 04-02 Table B Tr ansition of Fundamental FP Program a nd Design Elements (NFPA 805 Chapter 3)
Table B NFPA 805 Ch. 3 TransitionAmeren Missouri Callaway Plant NFPA 805 Transition ReportNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentHose stations protecting the ESW pump house are fed by the ESW system, not the fire protection water system. The NRC approved 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.NUREG-0830, "Safety Evaluation Report Related to the Operation of Callaway Plant, Unit No. 1," dated October 1981 / Section 9.5.1.6Submit for NRC Approval 3.6.1 For all power block buildings, Class III standpipe and hose systems shall be installed in accordance with NFPA 14, Standard for the Installation of Standpipe, Private Hydrant, and Hose Systems.Standpipe and hose station s comply with the requirements of this section, except for those protecting the ESW pump house as identified below.Calculation M-KC-452, "Hose Station Adequacy," Rev. 0 / AllCalculation KC-27, "NFPA Code Conformance Review," Rev. 0 / Appendix A, Section 24, Code Section 4-4.2Complies with Clarification 3.6.2 A capability shall be provided to ensure an adequate water flow rate and nozzle pressure for all hose stations. This capability includes the provision of hose station pressure reducers where necessary for the safety of plant industrial fire brigade members and off-site fire department personnel.Hose stations protecting the ESW pump house are fed by the ESW system, not the fire protection water system. The NRC approved 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.NUREG-0830, "Safety Evaluation Report Related to the Operation of Callaway Plant, Unit No. 1," dated October 1981 / Section 9.5.1.6Submit for NRC ApprovalA capability shall be provided to ensure an adequate water flow rate and nozzle pressure for all hose stations. This capability includes the provision of hose station pressure reducers where necessary for the safety of plant industrial fire brigade members and off-site fire department personnel.
August 2011 Page A-60 FPE RAI 07 Attachment A. NEI 04-02 Table B Tr ansition of Fundamental FP Program a nd Design Elements (NFPA 805 Chapter 3)
Table B NFPA 805 Ch. 3 TransitionAmeren Missouri Callaway Plant NFPA 805 Transition ReportNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentSections 2223 and 2231 of NFPA 72D-1975 Edition requires adequate secondary and remotely-located equipment power supplies. Page 9-3 of NUREG-0830 Supplement 3 states, "The SER states that the plant fire detection system is installed in accordance with NFPA 72D. During its site visit, the staff noted that the back-up power supply may not meet the recommendations of NFPA 72D. The applicant was unable to show compliance, and verbally agreed to prepare an analysis showing how the existing primary/back-up power supply circuitry compares to the requirements of NFPA 72D."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 Letter SLNRC 84-0014 from Petrick (SNUPPS) to Denton (NRC) dated February 1, 1984 / Enclosure 10NUREG-0830, "Safety Evaluation Report Related to the Operation of Callaway Plant, Unit No. 1," dated October 1981 / Section 9.5.1.6Calculation KC-27, "NFPA Code Conformance Review," Rev. 0 / Appendix A, Section 72DNFPA 72D, "Standard for the Installation, Maintenance, and Use of Proprietary Protective Signaling Systems for Watchman, Fire Alarm and Supervisory Service," 1975 Edition / Sections 1232, 2223, and 2231Complies by Previous NRC Approval3.8.1 Fire Alarm.Alarm initiating devices shall be installed in accordance with NFPA 72, National Fire Alarm Code. Alarm annunciation shall allow the proprietary alarm system to transmit fire-related alarms, supervisory signals, and trouble signals to the control room or other constantly attended location from which required notifications and response can be initiated. Personnel assigned to the proprietary alarm station shall be permitted to have other duties. The following fire-related signals shall be transmitted:
August 2011 Page A-65 LIC-05 LIC-05 Attachment A. NEI 04-02 Table B Tr ansition of Fundamental FP Program a nd Design Elements (NFPA 805 Chapter 3)
Table B NFPA 805 Ch. 3 TransitionAmeren Missouri Callaway Plant NFPA 805 Transition ReportNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentSee implementation item identified below.CAR 200902315, "NFPA 805 Transition - Site Organizations Support Tracking CAR"Modification MP 12-0009Complies, with Required Action3.9.1(1) NFPA 13, Standard for the Installation of Sprinkler SystemsIMPLEMENTATION ITEMS: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.11-805-091Modification MP 12-0009 will be completed to modify the quick-response sprinkler heads installed at an angle in cable chases to a configuration that is in accordance with the requirements of NFPA 13-1976 Edition.11-805-094Automatic and manual water 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.
N/A N/A3.9.1(2) NFPA 15, Standard for Water Spray Fixed Systems for Fire ProtectionWater mist fire protection systems are not used at Callaway Plant.
N/A N/A3.9.1(3) NFPA 750, Standard on Water Mist Fire Protection SystemsFoam-water sprinkler and foam-water spray systems are not used at Callaway Plant.
N/A N/A3.9.1(4) NFPA 16, Standard for the Installation of Foam-Water Sprinkler and Foam-Water Spray Systems August 2011 Page A-76 FPE RAI 14 FPE RAI 14 Attachment A. NEI 04-02 Table B Tr ansition of Fundamental FP Program a nd Design Elements (NFPA 805 Chapter 3)
Table B NFPA 805 Ch. 3 TransitionAmeren Missouri Callaway Plant NFPA 805 Transition ReportNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentM-22KC series drawings and Drawing J-1073-00052 identify that all waterflow alarms annunciate on panels that connect to KC008, which is located in the control room.FSAR SP, Section 9.5.1.2.2.1, Rev. OL-14f / Paragraph 3System Description 10466-M-00KC, "Fire Protection System Description," Rev. 4 / Section 3.1.3Drawing J-1073-00059, "KC008 and KC365 4120 Addressable Network Fire Alarm System Graphic Command Center Arrangement Details," Rev. 3 /
AllDrawing M-22KC01, "P&ID, Fire Protection Turbine Building," Rev. 21 /
AllDrawing M-22KC02, "P&ID, Fire Protection System Sheet 2," Rev. 21 /
AllDrawing M-22KC03, "P&ID, Fire Protection System Sheet 3," Rev. 24 /
AllDrawing M-22KC05, "P&ID, Fire Protection System Sheet 5," Rev. 11 /
AllDrawing M-22KC08, "P&ID, Fire Protection Preaction Sprinkler System Sheet 8," Rev. 11 / AllDrawing M-22KC09, "P&ID, Fire Protection System," Rev. 0 / AllDrawing J-1073-00052, "KC324 4120 Addressable Network Fire Alarm Control Panel System Operation Matrix," Rev. 4 / AllComplies 3.9.3 All alarms from fire suppression systems shall annunciate in the control room or other suitable constantly attended location.August 2011 Page A-78 FPE RAI 09 Attachment A. NEI 04-02 Table B Tr ansition of Fundamental FP Program a nd Design Elements (NFPA 805 Chapter 3)
Table B NFPA 805 Ch. 3 TransitionAmeren Missouri Callaway Plant NFPA 805 Transition ReportNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentClean agent fire extinguishing systems are not used at Callaway Plant.N/A N/A3.10.1(3) NFPA 2001, Standard on Clean Agent Fire Extinguishing SystemsM-22KC series drawings identify that all system actuation alarms annunciate on panels that connect to KC008, which is located in the control room.Drawing M-22KC04, "Fire Protection Halon System P&ID Sheet 4," Rev. 7 /
AllDrawing M-22KC06, "Fire Protection Halon System P&ID Sheet 6," Rev. 3Drawing M-22KC04, "Fire Protection Halon System P&ID Sheet 7," Rev. 7 /
AllDrawing J-1073-00059, "KC008 and KC365 4120 Addressable Network Fire Alarm System Graphic Command Center Arrangement Details," Rev. 3 /
AllComplies3.10.2 Operation of gaseous fire suppression systems shall annunciate and alarm in the control room or other constantly attended location identified.No Additional ClarificationCalculation KC-27, "NFPA Code Conformance Review," Rev. 0 / Appendix A, Section 12ACalculation KC-43, "NFPA 805 Code Comparison," Rev. 0 / Attachment 4Complies3.10.3 Ventilation system design shall take into account prevention from over-pressurization during agent injection, adequate sealing to prevent loss of agent, and confinement of radioactive contaminants.
August 2011 Page A-84 FPE RAI 09 Attachment A. NEI 04-02 Table B Tr ansition of Fundamental FP Program a nd Design Elements (NFPA 805 Chapter 3)
Table B NFPA 805 Ch. 3 TransitionAmeren Missouri Callaway Plant NFPA 805 Transition ReportNFPA 805 Ch. 3 Ref.Requirements/GuidanceCompliance StatementCompliance BasisReference DocumentCarbon dioxide extinguishing systems are not used in the Power Block.N/A N/A3.10.8 Positive mechanical means shall be provided to lock out total flooding carbon dioxide systems during work in the protected space.As identified in the NFPA Code Conformance Review of NFPA 12A, a full system discharge test was performed for all Halon systems as part of the initial acceptance testing. No thermal impacts were noted as a result of these system discharges.Calculation KC-27, "NFPA Code Conformance Review," Rev. 0 / Appendix A, Section 12AComplies with Clarification3.10.9 The possibility of secondary thermal shock (cooling) damage shall be considered during the design of any gaseous fire suppression system, but particularly with carbon dioxide.
August 2011 Page A-86 FPE RAI 13
 
Attachment B:  Revisions to Transition Report Attachment B - NEI 04-02
 
Table B Nuclear Safety
 
Capability Assessment -
 
Methodology Review 
 
Attachment B - NEI 04-02 TABLE B Nuclear Safety Capability Assessment Methodology ReviewAmeren MissouriCallaway Plant NFPA 805 Transition ReportNFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Reference Documents* 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 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 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 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.Calculation KC-26, "Nuclear Safety Capability Assessment," Revision 0NUREG-0830, SER Supplement No. 3, Docket No, STN 50-483, May 1984 NUREG-0830, SER Supplement No. 4, Docket No, STN 50-483, October 1984August 2011Page B-12 SSA RAI 03 Attachment B - NEI 04-02 TABLE B Nuclear Safety Capability Assessment Methodology ReviewAmeren MissouriCallaway Plant NFPA 805 Transition ReportNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Reference Documents 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 the subsequent deterministic fire area assessment for NFPA 805 NSCA and NPO components was performed utilizing the following criteria with respect to 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 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).Calculation KC-26, "Nuclear Safety Capability Assessment," Revision 0August 2011Page B-70 SSA RAI 02 Attachment B - NEI 04-02 TABLE B Nuclear Safety Capability Assessment Methodology ReviewAmeren MissouriCallaway Plant NFPA 805 Transition ReportNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis
* Spurious operation, when only resulting from positive to positive (+ to +) and negative to negative (- to -) external DC hot shorts in ungrounded DC circuits is only postulated in the circuit identification and analysis for high/low pressure interface valves and high consequence Fire PRA valves (as defined by the Fire PRA).* No credit is taken for self-healing of electrical failures.
* Multiple AC and DC grounds are postulated in the circuit identification and analysis. Multiple grounds in ungrounded AC or DC systems can result in clearing of fuses, or tripping of breakers.""* The circuit identification and analysis does not screen out cables on the basis of jacket material, insulation material, shielding, and/or the cable being routed in a dedicated conduit. However, the deterministic NSCA area-by-area analyses may discount spurious operation based on the fire affected cable being routed in a dedicated conduit, and therefore being protected from external sources of voltage."Circuit identification and analysis for the Callaway Plant NSCA does not include limiting assumptions as described in RIS 2004-03.Comments:The circuit analysis and cable selection performed for Callaway is consistent with the guidelines, criteria, and assumptions of NEI 00-01 Revision 2. However, it has become apparent that NEI 00-01 may be unclear to some individuals with respect to the guidance, criteria, and assumptions as pertaining to inter-cable hot shorts (i.e., direct inter-cable hot shorts - source cable to target cable, and indirect inter-cable hot shorts - source cable to target cable through a ground plane). As a consequence, Callaway is providing the following clarification in the NFPA 805 LAR to describe the Callaway circuit analysis and cable selection treatment for inter-cable hot shorts, inclusive of direct and indirect inter-cable hot shorts:The Callaway circuit analysis and cable selection process includes that a positive DC or a negative DC inter-cable hot short can occur on the same target cable so as to result in the spurious operation of a non-high/low pressure interface component.The Callaway circuit analysis and cable selection process excludes that a positive DC and a negative DC inter-cable hot short can occur on the same target cable so as to result in the spurious operation of a non-high/low pressure interface component.Inter-cable hot shorts are considered by Callaway to occur from direct source cable(s) to target cable interactions or from indirect source cable(s) to target cable interactions through a ground plane (i.e., the ground plane could be established through any fire affected plant equipment, conduits, and/or raceways). No distinction is made by Callaway between direct and indirect inter-cable hot shorts. The mechanism for the externally applied voltage source (i.e., hot short) to contact the target cable is treated as a "black box".Based on this treatment, a non-high/low pressure interface component cannot spuriously operate due to a single inter-cable hot short (positive DC or negative DC) so long as there are also no adequate sources of DC voltage originating within the target cable that could result in spurious operation of the non-high/low pressure interface component due to a combination of intra-cable short circuits and a single inter-cable hot short.August 2011Page B-114 LIC-17 Attachment B - NEI 04-02 TABLE B Nuclear Safety Capability Assessment Methodology ReviewAmeren MissouriCallaway Plant NFPA 805 Transition ReportNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Reference DocumentsCalculation KC-26, "Nuclear Safety Capability Assessment," Revision 0The Callaway treatment is consistent with NRC Generic Letter 86-10, Question and Answer 5.3.1, and NEI 00-01 Revision 2, Figure 3.5.2-5. The Callaway treatment is also consistent with the test results from NUREG/CR-7100, SAND2012-0323P, "Direct Current Electrical Shorting in Response to Exposure Fire (DESIREE-Fire): Test Results," specific to inter-cable hot shorts (Section 6.5.3). Note the test configuration for the inter-cable hot shorts from NUREG/CR-7100, SAND2012-0323P, as depicted in Figure A-54, was set up intentionally to obtain inter-cable hot shorts for the study, and is not representative of field typical installations which may further reduce the likelihood of inter-cable hot shorts.Multiple grounds (in ungrounded circuits) are considered in the Callaway circuit analysis and cable selection process with respect to the potential for loss of required power for ungrounded circuits. This approach is consistent with NEI 00-01 Revision 2, Figure 3.5.2-3.August 2011Page B-115 LIC-17 Attachment B - NEI 04-02 TABLE B Nuclear Safety Capability Assessment Methodology ReviewAmeren MissouriCallaway Plant NFPA 805 Transition ReportNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis This section provides guidance for analyzing the effects of a hot short on circuits for required safe shutdown equipment. A hot short is defined as a fire induced insulation breakdown between conductors of the same cable, a different cable or some other external source resulting in an undesired impressed voltage on a specific conductor. The potential effect of the undesired impressed voltage would be to cause equipment to operate or fail to operate in an undesired manner.Consider the following specific circuit failures related to hot shorts as part of the post-fire safe shutdown analysis:
- A hot short between an energized conductor and a de-energized conductor within the same cable may cause a spurious actuation of equipment. The spuriously actuated device (e.g., relay) may be interlocked with another circuit that causes the spurious actuation of other equipment. This type of hot short is called a conductor-to-conductor hot short or an internal hot short.- A hot short between any external energized source such as an energized conductor from another cable (thermoplastic cables only) and a de-energized conductor may also cause a spurious actuation of equipment. This is called a cable-to-cable hot short or an external hot short. Cable-to-cable hot shorts between thermoset cables are not postulated to occur pending additional research.NEI 00-01 RefNEI 00-01 Section 3 Guidance3.5.2.3Circuit Failures Due to a Hot ShortApplicableNoneApplicabilityCommentsAlignment StatementAlignment BasisAlignsCallaway Plant Calculation KC-26, Section 8.0, Circuit Identification and Analysis, identifies the overall process utilized to perform circuit identification and analysis for the NSCA components identified as being required to satisfy each of the Nuclear Safety Performance Criteria (NSPC) from Section 1.5.1 of NFPA 805.From Section 8.2 of KC-26:"e. Postulate the effects of open circuits, short circuits, and/or grounds upon the desired position(s) / function(s) for the component at-power and/or non-power, as applicable""* Multiple simultaneous circuit failures are postulated in the circuit identification and analysis (affecting multiple cables, affecting multiple conductors within cables). No limit is prescribed to the number or type circuit failures that are postulated to occur except as modified by the following: August 2011Page B-122 Attachment B - NEI 04-02 TABLE B Nuclear Safety Capability Assessment Methodology ReviewAmeren MissouriCallaway Plant NFPA 805 Transition ReportNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis
* Spurious operation, when resulting only from properly sequenced three-phase to three-phase external hot shorts is only postulated in the circuit identification and analysis for high/low pressure interface valves and high consequence Fire PRA valves (as defined by the Fire PRA).* Spurious operation, when only resulting from positive to positive (+ to +) and negative to negative (- to -) external DC hot shorts in ungrounded DC circuits is only postulated in the circuit identification and analysis for high/low pressure interface valves and high consequence Fire PRA valves (as defined by the Fire PRA).""* The circuit identification and analysis does not screen out cables on the basis of jacket material, insulation material, shielding, and/or the cable being routed in a dedicated conduit. However, the deterministic NSCA area-by-area analyses may discount spurious operation based on the fire affected cable being routed in a dedicated conduit, and therefore being protected from external sources of voltage."Comments:The circuit analysis and cable selection performed for Callaway is consistent with the guidelines, criteria, and assumptions of NEI 00-01 Revision 2. However, it has become apparent that NEI 00-01 may be unclear to some individuals with respect to the guidance, criteria, and assumptions as pertaining to inter-cable hot shorts (i.e., direct inter-cable hot shorts - source cable to target cable, and indirect inter-cable hot shorts - source cable to target cable through a ground plane). As a consequence, Callaway is providing the following clarification in the NFPA 805 LAR to describe the Callaway circuit analysis and cable selection treatment for inter-cable hot shorts, inclusive of direct and indirect inter-cable hot shorts:The Callaway circuit analysis and cable selection process includes that a positive DC or a negative DC inter-cable hot short can occur on the same target cable so as to result in the spurious operation of a non-high/low pressure interface component.The Callaway circuit analysis and cable selection process excludes that a positive DC and a negative DC inter-cable hot short can occur on the same target cable so as to result in the spurious operation of a non-high/low pressure interface component.Inter-cable hot shorts are considered by Callaway to occur from direct source cable(s) to target cable interactions or from indirect source cable(s) to target cable interactions through a ground plane (i.e., the ground plane could be established through any fire affected plant equipment, conduits, and/or raceways). No distinction is made by Callaway between direct and indirect inter-cable hot shorts. The mechanism for the externally applied voltage source (i.e., hot short) to contact the target cable is treated as a "black box".Based on this treatment, a non-high/low pressure interface component cannot spuriously operate due to a single inter-cable hot short (positive DC or negative DC) so long as there are also no adequate sources of DC voltage originating within the target cable that could result in spurious operation of the non-high/low pressure interface component due to a combination of intra-cable short circuits and a single inter-cable hot short.The Callaway treatment is consistent with NRC Generic Letter 86-10, Question and Answer 5.3.1, and NEI 00-01 Revision 2, Figure 3.5.2-5. The Callaway treatment is also consistent with the test results from NUREG/CR-7100, SAND2012-0323P, "Direct Current Electrical Shorting in Response to Exposure Fire (DESIREE-Fire): Test Results," specific to inter-cable hot shorts (Section 6.5.3). Note the test configuration for the inter-cable hot shorts from NUREG/CR-August 2011Page B-123 LIC-17 Attachment B - NEI 04-02 TABLE B Nuclear Safety Capability Assessment Methodology ReviewAmeren MissouriCallaway Plant NFPA 805 Transition ReportNFPA 805 Section: 2.4.2.2 Nuclear Safety Capability Circuit Analysis Reference DocumentsCalculation KC-26, "Nuclear Safety Capability Assessment," Revision 07100, SAND2012-0323P, as depicted in Figure A-54, was set up intentionally to obtain inter-cable hot shorts for the study, and is not representative of field typical installations which may further reduce the likelihood of inter-cable hot shorts.Multiple grounds (in ungrounded circuits) are considered in the Callaway circuit analysis and cable selection process with respect to the potential for loss of required power for ungrounded circuits. This approach is consistent with NEI 00-01 Revision 2, Figure 3.5.2-3.August 2011Page B-124 LIC-17
 
Attachment D:  Revisions to Transition Report Attachment D - Non-Power
 
Operational Modes Transition
 
SECURITY-RELATED INFORMATION - WITHHOLD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page D-6 SECURITY-RELATED INFORMATION - WITHHOLD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 FAQ 07-0040 Implementing GuidanceF.3 - Perform Fire Area Assessments (Identify pinch points)Identify locations where:  Fires may cause damage to the equipment (and cabling) credited above, or  KSFs are achieved solely by crediting recovery actions, e.g., alignment of gravity feed. Fire modeling may be used to determine if postulated fires in a fire area are expected to damage equipment (and cabling) thereby eliminating a pinch point. To implement this guidance perform the following tasks:  Determine if a single fire in the fire area can cause loss of success paths for a KSF. Conservatively, assume the entire contents of a fire area are lost. Document the loss of success paths. Specifically identify those areas that cause loss of all success paths for a KSF. If fire modeling is used to limit the damage in a fire area, document that fire modeling is credited and ensure the basis for acceptability of that model (location, type, and quantity of combustible, etc.) is documented. These critical design inputs should be maintained during outage modes. Fire modeling treatment should include an assessment of safety margin to account for uncertainties/accuracy of the fire model used.
ReviewA deterministic fire separation analysis (i.e., assuming full area burn) was performed as documented in Callaway Plant Calculation KC-26, "Nuclear Safety Capability Assessment," to identify pinch points (i.e., areas where redundant equipment and cables credited for a given
 
KSF fail due to fire damage). There is a total of eighty-one (81) fire areas at the Callaway Plant; however, for the purposes of performing the computerized deterministic NFPA 805 NSCA in SAFE-PB, Fire Areas A-16 and C-1 were each subdivided into two (2) unique analysis area IDs; and Fire Area RB-1 was subdivided into five (5) unique analysis area IDs. Consequently the deterministic NFPA 805 NSCA included the analysis of eighty-seven (87) fire areas. Fifty (50) fire areas were found to have an adequate number of KSF success paths to survive the entire contents loss of the fire area. Thirty-seven (37) fire areas were found to have pinch points resulting in the potential loss of one or more KSFs success paths. Fire modeling was not utilized to eliminate identification of pinch point fire areas as part of the implementation process for the step F.3 guidance from FAQ 07-0040. Callaway Plant aligns with FAQ 07-0040 implementing guidance, F.3, Perform Fire Area Assessments (Identify pinch points).
PRARAI04-f PRA RAI04-f Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page D-10 by fire area, the NPO risk reduction actions to be completed, and the evaluation and definition of the Callaway Plant POS's that are considered HREs. Fire Protection Program procedure APA-ZZ-00700, "Fire Protection Program," contains an overview of the NPO requirements, the commitments for implementation of the NPO risk reduction actions required by KC-26, "Nuclear Safety Capability Assessment," and a road map to identify the site specific implementing procedures used to implement the NPO requirements (Implementation Item 11-805-058). APA-ZZ-00741, "Control of Combustible Materials," contains controls to establish the outage roving fire watches that includes the required scope for the NPO risk reduction actions (Implementation Item 11-805-059). APA-ZZ-00742, "Control of Ignition Sources," contains controls to establish fire watches for the hot work activities including all plant operating states within the NPO scope. APA-ZZ-00703 , "Fire Protection Operability Criteria and Surveillance Requirements
,"contains the compensatory actions to be implemented should a fire protection system required to be operable during HRE periods be found to be impaired. In these cases continuous fire watches will be implemented in the affected systems areas (Implementation Item 11-805-061). EDP-ZZ-04044, "Fire Protection Reviews" contains guidance to ensure that changes to the fire protection program are reviewed for impact to the NPO requirements and risk reduction actions (Implementation Item 11-805-062). APA-ZZ-00315, "Configuration Risk Management," contains discussion on risk due to fire, NFPA 805 and the NPO requirements as part of risk management (Implementation Item 11-805-063). Guidance is contained within the outage control procedures to ensure that upon entry into the NPO plant operating states the outage roving fire watches are established. No specific requirements are necessary for the hot work controls because they are in place in all plant operating states. Additional guidance and controls are in place to ensure the HRE risk reduction tools are implemented prior to entry into a plant HRE. Guidance is also in place to monitor the plant state (T-Boil Times) to determine when the HRE is exited. OTN-BB-00002, "Reactor Coolant System Draining," contains a check list of the HRE risk reduction actions and requirements to implement the HRE risk reduction actions as a pre-requisite to conduct of the procedure which results in RCS reduced inventory states (Implementation Item 11-805-064). OTN-BB-00002, Addendum 7, "Raising RCS Level to 6 Inches Below the RX Vessel Flange," contains a decision point to check RCS level and T-Boil to determine if the HRE has been exited, then an HRE Risk Reduction Actions exit check list is provided (Implementation Item 11-805-065). OTN-BB-00001, "Reactor Coolant System," contains a requirement to maintain a continuous action to monitor T-Boil until the HRE can be exited then an HRE Risk Reduction Actions exit check list is provided (Implementation Item 11-805-066). OTG-ZZ-00007, "Refueling Preparation, Performance and Recovery," contains a requirement to maintain a continuous action to monitor T-Boil until the HRE can be exited then an HRE Risk Reduction Actions exit check list is provided (Implementation Item 11-805-067).
LIC-07
 
Attachment H:  Revisions to Transition Report Attachment H - NFPA 805
 
Frequently Asked Question
 
Summary Table 
 
Ameren Missouri Callaway Plant NFPA 805 Transition Report Table H NEI 04-02 FAQs Utilized in LAR Submittal No. Rev. Title FAQ Ref.
Closure Memo  08-0052 0 Transient Fire Growth Rate and Control Room Non-Suppression ML081500500 ML091590505 ML092120501 07-0054* 1 Demonstrating Compliance with Chapter 4 of NFPA 805 ML103510379 ML110140183 09-0056 2 Radioactive Release Transition ML102810600 ML102920405 08-0057 3 New Shutdown Strategy ML100330863 ML100960568 10-0059 5 NFPA 805 Monitoring ML120410589 ML120750108
  *Note: The FAQ Submittal number was 08-0054 but the NRC Closure Memo for the FAQ was listed as 07-0054. 07-0054 was used to be consistent with the Closure Memo.
This table includes FAQs that have not been approved by the NRC but are utilized in this submittal based on industry concurrence with the guidance contained therein:
MP RAI 01     
 
MP RAI 01   
 
Attachment J:  Revisions to Transition Report Attachment J - Fire Modeling
 
V&V 
 
Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page J-2Attachment J - Table J-1  -  V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion Flame Height  (Method of Heskestad) Calculates the vertical extension of the flame region of a fire. NUREG-1805, Chapter 3, 2004  NUREG-1824, Volume 3, 2007  SFPE Handbook, 4 th Edition, Chapter 2-1, Heskestad, 2008  The correlation is used in the NUREG-1805 fire model, for which V&V was documented in NUREG-1824. The correlation is documented in an authoritative publication of the "SFPE Handbook of Fire Protection Engineering." The correlation is applied within the validated range reported in NUREG-1824 or has been justified as acceptable by qualitative analysis or quantitative sensitivity analysis.
Plume Centerline Temperature (Method of Heskestad) Calculates the vertical separation distance, based on temperature, to a target in order to determine the vertical extent of the ZOI. NUREG-1805, Chapter 9, 2004  NUREG-1824, Volume 3, 2007  SFPE Handbook, 4 th Edition, Chapter 2-1, Heskestad, 2008  NUREG/CR-6850, Appendix H - Damage Criteria, 2005  The correlation is used in the NUREG-1805 fire model, for which V&V was documented in NUREG-1824. The correlation is documented in an authoritative publication of the "SFPE Handbook of Fire Protection Engineering."  The correlation is applied within the validated range reported in NUREG-1824 or has been justified as acceptable by qualitative analysis or quantitative sensitivity analysis. NUREG/CR-6850 generic screening damage criteria is used, which is considered conservative.
LIC-10 LIC-10 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page J-3Attachment J - Table J-1  -  V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion Radiant Heat Flux (Point Source Method) Calculates the horizontal separation distance, based on heat flux, to a target in order to determine the horizontal extent of the ZOI. NUREG-1805, Chapter 5. 2004  NUREG-1824, , Volume 4, 2007  SFPE Handbook, 4 thedition, Chapter 3-10, Beyler, C., 2008  NUREG/CR-6850, Appendix H - Damage Criteria, 2005  The correlation is used in the NUREG-1805 fire model, for which V&V was documented in NUREG-1824. The correlation is documented in an authoritative publication of the "SFPE Handbook of Fire Protection Engineering."  The correlation is applied within the validated range reported in NUREG-1824 or has been justified as acceptable by qualitative analysis or quantitative sensitivity analysis. NUREG/CR-6850 generic screening damage criteria is used, which is considered conservative. Plume Radius (Method of Heskestad) Calculates the horizontal radius, based on temperature, of the plume at a given height.
The correlation is derived of the Heskestad centerline plume correlation. FIVE-Rev1, Referenced by EPRI Report 1002981, 2002  SFPE Handbook, 4 thEdition, Chapter 2-1, Heskestad, G., 2008  NUREG/CR-6850, Appendix H - Damage Criteria, 2005  The correlation is used in the FIVE-Rev1 fire model. The correlation is documented in an authoritative publication of the "SFPE Handbook of Fire Protection Engineering." The Heskestad centerline plume correlation V&V is documented in NUREG-1824. The Heskestad correlation is applied within the validated range reported in NUREG-1824 or has been justified as acceptable by qualitative analysis or quantitative sensitivity analysis. NUREG/CR-6850 generic screening damage criteria is used, which is considered conservative.
LIC-10 LIC-10 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page J-4Attachment J - Table J-1  -  V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion Hot Gas Layer (Method of MQH) Calculates the hot gas layer temperature for a room with natural ventilation. NUREG-1805, Chapter 2, 2004  NUREG-1824, Volume 3, 2007  SFPE Handbook, 4 th Edition, Chapter 3-6, Walton W. and Thomas, P., 2008  The correlation is used in the NUREG-1805 fire model, for which V&V was documented in NUREG-1824. The correlation is documented in an authoritative publication of the "SFPE Handbook of Fire Protection Engineering." The correlation is applied within the validated range reported in NUREG-1824 or has been justified as acceptable by qualitative analysis or quantitative sensitivity analysis.Hot Gas Layer (Method of Beyler) Calculates the hot gas layer temperature for a closed compartment with no ventilation. NUREG-1805, Chapter 2, 2004  NUREG-1824, Volume 3, 2007  SFPE Handbook, 4 th Edition, Chapter 3-6, Walton W. and Thomas, P., 2008  The correlation is used in the NUREG-1805 fire model, for which V&V was documented in NUREG-1824. The correlation is documented in an authoritative publication of the "SFPE Handbook of Fire Protection Engineering."The correlation is applied within the validated range reported in NUREG-1824 or has been justified as acceptable by qualitative analysis or quantitative sensitivity analysis.
LIC-10 LIC-10 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page J-5Attachment J - Table J-1  -  V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion Hot Gas Layer (Method of Foote, Pagni, and Alvares
[FPA])Calculates the hot gas layer temperature for a room with forced ventilation. NUREG-1805, Chapter 2, 2004  NUREG-1824, Volume 3, 2007  SFPE Handbook, 4 th Edition, Chapter 3-6, Walton W. and Thomas, P., 2008  The correlation is used in the NUREG-1805 fire model, for which V&V was documented in NUREG-1824. The correlation is documented in an authoritative publication of the "SFPE Handbook of Fire Protection Engineering."The correlation is applied within the validated range reported in NUREG-1824 or has been justified as acceptable by qualitative analysis or quantitative sensitivity analysis. Hot Gas Layer (Method of Deal and Beyler)Calculates the hot gas layer temperature for a room with forced ventilation. NUREG-1805, Chapter 2, 2004  NUREG-1824, Volume 3, 2007  SFPE Handbook, 4 th Edition, Chapter 3-6, Walton W. and Thomas, P.,
2008  The correlation is used in the NUREG-1805 fire model, for which V&V was documented in NUREG-1824. The correlation is documented in an authoritative publication of the "SFPE Handbook of Fire Protection Engineering." The correlation is applied within the validated range reported in NUREG-1824 or has been justified as acceptable by qualitative analysis or quantitative sensitivity analysis.
LIC-10 LIC-10 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page J-6Attachment J - Table J-1  -  V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion Ceiling Jet Temperature (Method of Alpert) Calculates the horizontal separation distance, based on temperature at the ceiling of a room, to a target in order to determine the horizontal extent of the ZOI. FIVE-Rev1, Referenced by EPRI Report 1002981, 2002  NUREG-1824, Volume 4, 2007  SFPE Handbook, 4 th Edition, Chapter 2-2, Alpert, R., 2008  NUREG/CR-6850, Appendix H - Damage Criteria, 2005  The correlation is used in the FIVE-Rev1 fire model, for which V&V was documented in NUREG-1824. The correlation is documented in an authoritative publication of the "SFPE Handbook of Fire Protection Engineering."  The correlation is applied within the validated range reported in NUREG-1824 or has been justified as acceptable by qualitative analysis or quantitative sensitivity analysis. NUREG/CR-6850 generic screening damage criteria is used, which is considered conservative. Hot Gas Layer Calculations using Fire Dynamics Simulator (Version 5) Used to calculate the hot gas layer temperatures for various compartments, and the layer height. FDS Version 5  NIST Special Publication 1018-5, Volume 2:
"Verification" NIST Special Publication 1018-5, Volume 3:
 
"Validation"  NUREG-1824, Volume 7, 2007  V&V of the FDS is documented in NIST Special Publication 1018-5. The V&V of FDS specifically for Nuclear Power Plant applications has also been documented in NUREG-1824. The models are applied within their validated range reported in NUREG-1824  or have been justified as acceptable by qualitative analysis or quantitative sensitivity analysis.
LIC-10 LIC-10 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page J-7Attachment J - Table J-1  -  V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion Sprinkler Actuation Calculation using Fire Dynamics Simulator (Version 5)
Used to estimate sprinkler actuation timing based on ceiling jet temperature, velocity, and thermal response of sprinkler. FDS Version 5  NIST Special Publication 1018-5, Volume 2:
 
"Verification" NIST Special Publication 1018-5, Volume 3:
 
"Validation"  NUREG-1824, Volume 7, 2007  V&V of the FDS is documented in NIST Special Publication 1018-5. The V&V of FDS (for ceiling jet temperature) specifically for Nuclear Power Plant applications has also been documented in NUREG-1824. The models are applied within their validated range reported in NUREG-1824  or have been justified as acceptable by qualitative analysis or quantitative sensitivity analysis Smoke Detection Actuation Correlation (Method of Heskestad and Delichatsios)
Alpert Ceiling Jet used to determine temperature and Heskestad and Delichatsios temperature to smoke density for smoke detection timing estimates. NUREG-1805, Chapter 11, 2004  NUREG-1824, Volume 4, 2007  SFPE Handbook, 4 th Edition, Chapter 4-1, Custer R., Meacham B., and Schifiliti, R., 2008  SFPE Handbook, 4 th Edition, Chapter 2-2, Alpert, R., 2008  The smoke detection correlation is used in the NUREG-1805 fire model. Alpert's ceiling jet correlation V&V is documented in NUREG-1824. The correlation is applied within the validated range reported in NUREG-1824 or has been justified as acceptable by qualitative analysis or quantitative sensitivity analysis. The temperature to smoke density correlation is documented in an authoritative publication of the "SFPE Handbook of Fire Protection Engineering."
LIC-10 LIC-10 FM RAI 01-f FM RAI 01-c Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page J-8Attachment J - Table J-1  -  V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion Sprinkler Activation Correlation Used to estimate sprinkler actuation timing based on the Alpert ceiling jet temperature, velocity, and thermal response of sprinkler. NUREG-1805, Chapter 10, 2004  NFPA Handbook, 19 th Edition, Chapter 3-9, Budnick, E., Evans, D., and Nelson, H., 2003  The sprinkler actuation correlation is used in the NUREG-1805 fire model. The correlation is documented in an authoritative publication of the NFPA Fire Protection Handbook. Alpert's ceiling jet correlation V&V is documented in NUREG-1824
.The correlation is applied within the validated range reported in NUREG-1824 or has been justified as acceptable by qualitative analysis or quantitative sensitivity analysis.
Control Room Abandonment Calculation using CFAST Evaluates the time at which control room abandonment is necessary based on smoke obscuration and average HGL temperature. NIST Special Publication 1086, 2008  CFAST Version 6  NUREG-1824, Volume 6, 2007  NUREG/CR-6850, Appendix H - Damage Criteria, 2005  V&V of the CFAST code is documented in the NIST Special Publication 1086. The V&V of CFAST specifically for Nuclear Power Plant applications has also been documented in NUREG-1824. The models are applied within their validated range reported in NUREG-1824 or have been justified as acceptable by qualitative analysis or quantitative sensitivity analysis. NUREG/CR-6850 generic screening damage criteria is used, which is considered conservative.
LIC-10 LIC-10 LIC-10 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page J-9Attachment J - Table J-1  -  V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion Temperature Sensitive Equipment Hot Gas Layer Study Determine the upper and lower gas layer temperatures for various compartments, and the layer height, for use in assessing damage to temperature sensitive equipment. NIST Special Publication 1086, 2008  CFAST Version 6  NUREG-1824, Volume 6, 2007  NUREG/CR-6850, Appendix H - "Damage Criteria, 2005"  V&V of the CFAST code is documented in the NIST Special Publication 1086. The V&V of CFAST specifically for Nuclear Power Plant applications has also been documented in NUREG-1824. The models are applied within their validated range reported in NUREG-1824 or have been justified as acceptable by qualitative analysis or quantitative sensitivity analysis. NUREG/CR-6850 generic screening damage criteria is used, which is considered conservative. Temperature Sensitive Equipment Zone of Influence Study Determine the radiant heat flux ZOI at which temperature sensitive equipment will reach damage thresholds. FDS Version 5  NIST Special Publication 1018-5, Volume 2:
 
"Verification" NIST Special Publication 1018-5, Volume 3:
 
"Validation"  NUREG-1824, Volume 7, 2007  NUREG/CR-6850, Appendix H - "Damage Criteria, 2005"  V&V of the FDS is documented in the NIST Special Publication 1018-5. The V&V of FDS specifically for Nuclear Power Plant applications has also been documented in NUREG-1824. The models are applied within their validated range reported in NUREG-1824  or have been justified as acceptable by qualitative analysis or quantitative sensitivity analysis. NUREG/CR-6850 generic screening damage criteria is used, which is considered conservative.
LIC-10 LIC-10 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page J-10Attachment J - Table J-1  -  V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion Plume/Hot Gas Layer Interaction Study Determine the point at which hot gas layer and plume interact and establish limits for plume temperature application. FDS Version 5  NIST Special Publication 1018-5, Volume 2:
 
"Verification" NIST Special Publication 1018-5, Volume 3:
 
"Validation"  NUREG-1824, Volume 7, 2007  NUREG/CR-6850, Appendix H - "Damage Criteria, 2005"  V&V of the FDS is documented in NIST Special Publication 1018-5. The V&V of FDS specifically for Nuclear Power Plant applications has also been documented in NUREG-1824. The models are applied within their validated range reported in NUREG-1824  or have been justified as acceptable by qualitative analysis or quantitative sensitivity analysis. NUREG/CR-6850 generic screening damage criteria is used, which is considered conservative.
LIC-10 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page J-11Attachment J - Table J-1  -  V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion Corner and Wall HRR Determines a heat release rate adjustment factor for fires that are proximate to a wall or corner. Zukoski, E.E., "Properties of Fire Plumes," Combustion Fundamentals of Fire, Cox, G., Ed., Academic Press, London, 1995  Sargent, W.S., "Natural Convection Flows and Associated Heat Transfer Processes in Room Fires,"Ph.D. thesis, California Institute of Technology, Pasadena, CA 1983  Cetegen, B.M., "Entrainment and Flame Geometry of Fire Plumes," Ph.D. thesis, California Institute of Technology, Pasadena, CA, 1982  Williamson, R.B. Revenaugh, A. and Mowrer, F.W., "Ignition Sources in Room Fire Tests and Some Implications for Flame Spread Evaluation," International Association of Fire Safety Science, Proceedings of the Third International Symposium, New York, pp. 657-666, 1991The correlation is applied within the validated range reported in the referenced studies or has been justified as acceptable by qualitative analysis or quantitative sensitivity analysis.
LIC-10 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page J-12Attachment J - Table J-1  -  V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion Correlation for Heat Release Rates of Cables  (Method of Lee) Used to correlate bench-scale data to heat release rates from cable tray fires. NUREG/CR-6850, Appendix R, 2005  SFPE Handbook, 4 th Edition, Chapter 3-1, Babrauskas, 2008 NBISR 85-3195, July 1985 The correlation is recommended by NUREG/CR-6850. The correlation is documented in an authoritative publication of the "SFPE Handbook of Fire Protection Engineering." The correlation is applied to configurations similar to those reported in NBISR 85-3195 or has been justified as acceptable by qualitative analysis.Correlation for Flame Spread over Horizontal Cable Trays (FLASH-CAT) Used to predict the growth and spread of a fire within a vertical stack of horizontal cable trays. NUREG/CR-7010, Section 9, 2010  NUREG/CR-6850, Appendix R, 2005  The correlation is recommended by NUREG/CR-7010 and follows guidance set forth in NUREG/CR-6850. The FLASH-CAT model is validated in NUREG/CR-7010, Section 9.2.3, through experimentally measured HRRs compared with the predictions of the FLASH-CAT model. The model is applied to configurations similar to those reported NUREG/CR-7010or has been justified as acceptable by qualitative analysis.
LIC-10 LIC-10 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page J-13Attachment J - Table J-1  -  V & V Basis for Fire Models / Model Correlations Used Calculation Application V & V Basis Discussion Smoke Detector Activation using Fire Dynamics Simulator (Version 5) Used to predict detector activation based on smoke production and velocity, as well as detector geometry and optical response. FDS Version 5  NIST Special Publication 1018-5, Volume 2:
"Verification"NIST Special Publication 1018-5, Volume 3: "Validation" Section 2.2.3 R. Roby, et al. "A Smoke Detector Algorithm for Large Eddy Simulation Modeling," National Institute of Standards and Technology, Gaithersburg, Maryland, July 2007. NIST GCR 07-911  T. Cleary, et al. "Fire Detector Performance Predictions in a Simulated Multi-Room Configuration." In Proceedings of the 12th International Conference on Automatic Fire Detection (AUBE '01). National Institute of Standards and Technology, Gaithersburg, Maryland, March 2001.
NIST SP 965. 12  V&V of the FDS model is documented in NIST Special Publication 1018-5. Roby, et al. validated the smoke detector algorithm against a number of fire scenarios and geometries and concluded that the algorithm and FDS model are accurately predicting the activation times of the smoke detectors. Cleary, et al. concluded that multi-room fire simulation with the FDS software can yield environmental conditions a detector or sensor may experience during an actual fire. The model is applied to configurations similar to those reported in Roby, et al. or has been justified as acceptable by qualitative analysis or quantitative sensitivity analysis.
FM RAI 01-c Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page J-14 Table J-1
 
==References:==
: 1. NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities," U.S. Nuclear Regulatory Commission, Washing ton,DC, September 2005. 2. NUREG/CR-7010, "Cable Heat Release, Ignition, and Spread in Tray Installations During Fire (CHRISTIFIRE), Volume 1: Horizontal Trays", Draft Report for Comment, United States Nuclear Regulatory Commission, October, 2010. 3. "The SFPE Handbook of Fire Protection Engineering," 4 th Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2008. 4. "The NFPA Fire Protection Handbook," 19 th Edition, A. E. Cote, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2003. 5. Peacock, R.D., Jones, W.W., Reneke, P.A., and Forney, G.P., "CF AST - Consolidated Model of Fire Growth and Smoke Transport (Version6) User's Guide," NIST Special Publication 1041, National Institute of Standards and Technology, Gaithersburg, MD, December 200 5.6. NIST Special Publication 1086, "CFAST - Consolidated Model of Fire Growth and Smoke Transport (Version 6) Software Development and Model Evaluation Guide", National Institute of Standards and Technology, Gaithersburg, MD, December 2008. 7. NIST Special Publication 1018-5, "Fire Dynamics Simulator (Version 5) Technical Reference Guide, Volume 2: Verification," Na tional Institute of Standards and Technology, October 29, 2010 8. NIST Special Publication 1018-5, "Fire Dynamics Simulator (Version 5) Technical Reference Guide, Volume 3: Validation," National Institute of Standards and Technology, October 29, 2010 9. "Fire Modeling Guide for Nuclear Power Plant Applications," EPRI 1002981, FINAL REPORT, August 2002.
10.Inspection M anual Chapter (IMC) 0609, Appendix F, "Fire Protection Significance Determination Process," Issue Date 02/28/05.11. R. Roby, et.al. "A Smoke Detector Algorithm for Large Eddy Simulation Modeling, National Institute of Standards and Technol ogy, Gaithersburg," Maryland, July 2007. NIST GCR 07-911. 12. T. Cleary, et.al. "Fire Detector Performance Predictions in a Simulated Multi-Room Configuration." In Proceedings of the 12 th International Conference on Automatic Fire Detection (AUBE '01). National Institute of Standards and Technology, Gaithersburg, Maryland, March 2001. NIST SP 965. 12. 13. Lee, B.T., NBISR 85-3195, "Heat Release Rate Characteristics of Some Combustible Fuel Sources in Nuclear Power Plants," Jul y 1985. LIC-10 FM RAI -01i LIC-10
 
Attachment L:  Revisions to Transition Report Attachment L - NFPA 805
 
Chapter 3 Requirements for
 
Approval (10 CFR
 
50.48(c)(2)(vii)) 
 
Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page L-2 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:  Shift Manager/CRS approval is obtained and documented. A Fire Protection Impairment is generated to document the approvals, intended usage and administrative controls in place using the fire protection impairment program (FPIP). Both fire water storage tanks are functional and have sufficient tank level margin based on the anticipated usage to remain functional during usage. Fire Water storage tank water level will be monitored to ensure the fire water storage tank's level remains above 260,000 gallons during use.
FPERAI11 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page L-3  Controls/communications are in place to ensure the non-fire protection system water demand can be secured immediately if a fire occurs. 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:  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. Monitoring of fire water storage tank levels ensures the two tanks' water volume will be maintained above the procedurally required limit of 260,000 gallons.
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 above controls adequate water flow will be available for the manual fire suppression 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.
FPERAI11 FPE RAI11 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page L-4
 
== Conclusion:==
NRC approval is requested for approval of the temporary use of the fire protection water supply with the following restrictions:  Shift manager/CRS approval is obtained and documented;  A Fire Protection Impairment is generated to document the approvals, intended usage and administrative controls in place using the fire protection impairment program (FPIP). Both fire water storage tanks are functional and have sufficient tank level margin based on the anticipated usage to remain functional during usage. Fire Water storage tank water level will be monitored to ensure the fire water storage tank's level remains above 260,000 gallons during use.
Controls/communications are in place to ensure the non-fire protection water demand can be secured immediately if a fire occurs;  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).
FPERAI11 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page L-8 Approval Request 3 In accordance with 10 CFR 50.48(c)(2)(vii), "Performance based methods," the fire protection program elements and minimum design requirements of Chapter 3 may be subject to the performance-based methods permitted elsewhere in the standard. In accordance with NFPA 805 Section 2.2.8, the performance-based approach to satisfy the nuclear safety, radiation release, life safety, and property damage/business interruption performance criteria requires engineering analyses to evaluate whether the performance criteria are satisfied. In accordance with 10 CFR 50.48(c)(2)(vii), the engineering analysis performed shall determine that the performance-based approach utilized to evaluate a variance from the requirements of NFPA 805 Chapter 3: (A) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; (B)Maintains safety margins; and (C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire nuclear safety capability). Ameren Missouri requests formal approval of performance based exceptions to the requirements in Chapter 3 of NFPA 805 as follows: NFPA 805, Section 3.2.3(1) NFPA 805, Section 3.2.3(1) states: "Procedures shall be established for implementation of the fire protection program. In addition to procedures that could be required by other sections of the standard, the procedures to accomplish the following shall be established: Inspection, testing, and maintenance for fire protection systems and features credited by the fire protection program."  Callaway Plant will utilize performance based methods to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features required by NFPA 805. Performance-based inspection, testing, and maintenance frequencies will be established as described in Electric Power Research Institute (EPRI) Technical Report TR-1006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features", Final Report, July 2003.
Basis for Request:
NFPA 805 Section 2.6, Monitoring, requires that "A monitoring program shall be established to ensure that the availability and reliability of the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria. Monitoring shall ensure that the assumptions in the engineering analysis remain valid."    NFPA 805 Section 2.6.1, Availability, Reliability, and Performance Levels, requires that "Acceptable levels of availability, reliability, and performance shall be established."NFPA 805 Section 2.6.2 requires that "Methods to monitor availability, reliability, and performance shall be established. The methods shall consider the plant operating experience and industry operating experience."
LIC-15 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page L-9 The scope and frequency of the inspection, testing, and maintenance activities for fire protection systems and features required in the fire protection program have been established based on the previously approved Technical Specifications / License Controlled Documents and appropriate NFPA codes. This request does not involve the use of the EPRI Technical Report TR-1006756 to establish the scope of those activities as that is determined by the required systems review identified in Table 4-3. This request is specific to the use of EPRI Technical Report TR-1006756 to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features credited by the fire protection program. As stated in EPRI Technical Report TR-1006756 Section 10.1, "The goal of a performance-based surveillance program is to adjust test and inspection frequencies commensurate with equipment performance and desired reliability."  This goal is consistent with the stated requirements of NFPA 805 Section 2.6. The EPRI Technical Report TR-1006756 provides an accepted method to establish appropriate inspection, testing, and maintenance frequencies which ensure the required NFPA 805 availability, reliability, and performance goals are maintained. The target tests, inspections and maintenance will be those activities for the NFPA 805 required Fire Protection systems and features. The reliability and frequency goals will be established to ensure the assumptions in the NFPA 805 engineering analysis remain valid. The failure criterion will be established based on the required Fire Protection systems and features credited functions and will ensure those functions are maintained. Data collection and analysis will follow the Technical Report TR-1006756 document guidance. The failure probability will be determined based on the Technical Report TR-1006756 guidance and a 95% confidence level will be utilized. The performance monitoring will be performed in conjunction with the Monitoring program required by NFPA 805 section 2.6 and it will ensure site specific operating experience is considered in the monitoring process. The following is a flow chart that identifies the basic process that will be utilized.
LIC-15 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page L-10 EPRI TR-1006756 - Figure 10-1 Flowchart for Performance-Based Surveillance Program LIC-15 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page L-11 Acceptance Criteria Evaluation: Nuclear Safety and Radiological Release Performance Criteria: Use of performance based test frequencies established per EPRI Technical Report TR-1006756 methods combined with NFPA 805 Section 2.6, Monitoring Program, will ensure that the availability and reliability of the fire protection systems and features are maintained to the levels assumed in the NFPA 805 engineering analysis. Therefore, there is no adverse impact to Nuclear Safety Performance Criteria by the use of the performance based methods in EPRI Technical Report TR-1006756.      The radiological release performance criteria are satisfied based on the determination of limiting radioactive release (Refer to Attachment E of this LAR). FP Systems and features are credited as part of that evaluation. Use of performance based test frequencies established per EPRI Technical Report TR-1006756 methods combined with NFPA 805 Section 2.6, Monitoring Program will ensure that the availability and reliability of the fire protection systems and features are maintained to the levels assumed in the NFPA 805 engineering analysis which includes those assumptions credited to meet the Radioactive Release performance criteria. Therefore, there is no adverse impact to Radioactive Release performance criteria. Safety Margin and Defense-in-Depth: Use of performance based test frequencies established per EPRI Technical Report TR-1006756 methods combined with NFPA 805 Section 2.6, Monitoring Program will ensure that the availability and reliability of the fire protection systems and features are maintained to the levels assumed in the NFPA 805 engineering analysis which includes those assumptions credited in the Risk Evaluation safety margin discussions. In addition, the use of these methods in no way invalidates the inherent safety margins contained in the codes used for design and maintenance of fire protection systems and features. Therefore, the safety margin inherent and credited in the analysis has been preserved. The three echelons of defense-in-depth described in NFPA 805 section 1.2 are 1) to prevent fires from starting (combustible/hot work controls), 2) rapidly detect, control and extinguish fires that do occur thereby limiting damage (fire detection systems, automatic fire suppression, manual fire suppression, pre-fire plans), and 3) provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed (fire barriers, fire rated cable, success path remains free of fire damage, recovery actions). Echelon 1 is not affected by the use of EPRI Technical Report TR-1006756 methods. Use of performance based test frequencies established per EPRI Technical Report TR-1006756 methods combined with NFPA 805 Section 2.6 Monitoring Program, will ensure that the availability and reliability of the fire protection systems and features credited for DID are maintained to the levels assumed in the NFPA 805 engineering analysis. Therefore, there is no adverse impact to echelons 2 and 3 for the defense in depth.
LIC-15 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page L-12
 
==
Conclusion:==
NRC approval is requested for use of the performance based methods contained in Electric Power Research Institute (EPRI) Technical Report TR-1006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features", Final Report, July 2003 to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features required by NFPA 805. As described above, this approach is considered acceptable because it:  (A) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; (B) Maintains safety margins; and (C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire nuclear safety capability). 
 
Attachment S:  Revisions to Transition Report Attachment S - Plant
 
Modifications and Items to be
 
completed during
 
Implementation
 
Ameren Missouri  Callaway Plant NFPA 805 Transition Report August 2011 Page S-2  Tables S-1 and S-2, Plant Modifications Completed and Committed, respectively. Each of these tables provided below includes a description of the modifications along with the following information:  A problem statement,  Risk ranking of the modification based on estimated impact on the Fire PRA results (see legend),  An indication if the modification is currently included in the FPRA,  Compensatory Measure in place, and  A risk-informed characterization of the modification and compensatory measure. The following legend should be used when reviewing the tables:
o High = Modification would have an appreciable impact on reducing overall fire CDF.
o Medium = Modification would have a measurable impact on reducing overall fire CDF.
o Low = Modification would have either an insignificant or no impact on reducing overall fire CDF.
Table S Plant Modifications Completed Item Rank Problem Statement Proposed Modification In FPRAComp Measure Risk Informed Characterization 07-0066 M Buried carbon steel ESW system piping needed replacement. As part of this piping modification, relocate cables currently in nonconformance with 20 foot
 
separation criteria.
The buried carbon steel ESW piping was replaced with high density polyethylene (HDPE) piping. During the piping replacement the cabling associated with  EFTE0067A and  68A was relocated to restore the required 20 foot
 
separation criteria. Y No Cables affect ESW cooling from the UHS cooling towers, potentially failing both trains, but could be mitigated by a recovery action (and the recovery action is no longer needed). This is judged to have a medium impact on risk.
 
Compensatory measure:
None; modification complete Ameren Missouri  Callaway Plant NFPA 805 Transition Report August 2011 Page S-3  Table S Plant Modifications Completed Item Rank Problem Statement Proposed Modification In FPRAComp Measure Risk Informed Characterization 10-0032 H Risk metrics indicated that additional defense in depth was warranted for the AFW system.
This modification provides margin for AFW system MSPI metrics. Installed a non-safety related AFW pump as diverse AFW backup supply to the safety related motor driven and turbine driven pumps.
Y 4 No Fire PRA credits this modification for decay heat removal redundancy.
 
Compensatory measure:
None; modification complete 10-0038 H Improve Callaway Plant's defense in depth to mitigate the
 
consequences from a potential Station Black Out (SBO). Provide an alternate emergency source of power that is diverse from the Emergency Diesel Generators and offsite sources. Install four non-safety related diesel generators (8 MW) at the electric cooperative substation. Either the electric cooperative substation or the 4 non-safety diesel generators will be able to power either Safety Related bus in the event of a loss of AC
 
power and failure of the Emergency Diesel Generators.
Y 4 No Fire PRA credits this modification for electrical power redundancy 
 
Compensatory measure:
None; modification installed. Additional changes forthcoming that do not affect FPRA.
 
Table S-2, Items provided below are those modifications that will be completed prior to the implementation of the new NFPA 805 FP program. Currently open modifications will be field completed no later than June 30, 2013. Appropriate compensatory measures for any inc omplete NFPA 805 related modifications will be maintained until the modifications are complete.
4 Refer to associated implementation item in Table S-3.
FPE RAI 14 Ameren Missouri  Callaway Plant NFPA 805 Transition Report August 2011 Page S-4  Table S Plant Modifications Committed Item Rank Problem Statement Proposed Modification In FPRAComp Measure Risk Informed Characterization 05-3029 - Install lower amperage fuses to prevent damage to 14 AWG cables and prevent secondary fires from occurring in the MCR.
Install lower amperage fuses for various 14 AWG control circuits in the MCR. The majority of the modification centers around the trip circuit fuses on NB, NG, PA, PB, and PG system breakers.
N No This modification ensures there are no secondary fires. NUREG/CR-6850 methodology does not address secondary fires, but the issue of secondary fires was raised during the pilot plant RAI process.
Secondary fires are not modeled in the Callaway Fire PRA and an assessment of risk was not performed.
Compensatory measure for NFPA 805: In accordance with station procedures, appropriate compensatory measures will be established when the NFPA 805 fire
 
protection program becomes effective and remain in effect until this modification is complete.
Compensatory measure for Current Fire Protection Licensing Basis:
None; the MCR is deterministically compliant with the Current Fire Protection Licensing Basis.
Ameren Missouri  Callaway Plant NFPA 805 Transition Report August 2011 Page S-5  Table S Plant Modifications Committed Item Rank Problem Statement Proposed Modification In FPRAComp Measure Risk Informed Characterization 07-0151 L During a fire in the main control room (MCR), selected cables to B train related equipment fed from NB02 will be isolated to prevent a multi-spurious hot short from stopping or starting safety equipment. Circuits that have isolation switches but lack redundant fuses are included in this modification. This modification will eliminate credit previously taken to have operators replace potentially blown fuses prior to the NFPA 805 transition. Install redundant fuses and isolation switches for MCR evacuation procedure OTO-ZZ-00001. Y Y FPIP 14050 PRA assumes that after a fire in the main control room, the B train components are operable from the auxiliary shutdown panel without requiring replacement of fuses. Compensatory measure for NFPA 805: In accordance with station procedures, appropriate compensatory measures will be established when the NFPA 805 fire
 
protection program becomes effective and remain in effect until this modification is complete.
Compensatory measure for Current Fire Protection Licensing Basis:
None; the MCR is deterministically compliant with the Current Fire Protection Licensing Basis.
Ameren Missouri Callaway Plant NFPA 805 Transition Report Table S Plant Modifications Committed Item Rank Problem Statement Proposed Modification In FPRA Comp Measure  Risk Informed Characterization 09-0025 M A fire in fire areas A-1, A-2, A-4, A-8, A-16 (analysis area A-16S), A-27, C-18, C-21, C-22, C-23, C-24, C-30, C-33, and RB-1 (analysis areas RB1, RB2, RB3, and RB4) could cause EJHV8811A and/or B to spuriously open due to direct valve control cable damage and begin draining the RWST to the containment emergency sumps. To protect against multiple spurious scenarios, the solution is to run a single wire in a protected metal jacket such that spurious valve opening due to a hot short affecting the valve control circuit is eliminated for these fire areas.
Y Y FPIP 14050 This is judged to be a moderate risk improvement. In A-1, A-2, A-4, A-8, A-16 (analysis area A-16S), A-27, C-18, C-21, C-22, C-23, C-24, C-30, C-33, and RB-1 (analysis areas RB1, RB2, RB3, and RB4), EJHV8811A/B are assumed to not have potential for spurious opening due to valve control cable damage because of the modification.
EJHV8811A/B can still spuriously open if the MCC which powers the valve is involved in the fire, or in response to a valid or spurious SI signal concurrent with a spurious RWST Low level signal.
Compensatory measure for NFPA 805: In accordance with station procedures, appropriate compensatory measures will be established when the NFPA 805 fire protection program becomes effective and remain in effect until this modification is complete.
Compensatory measure for Current Fire Protection Licensing Basis : None; fire areas A-1, A-2, A-4, A-8, A-16 (analysis area A-16S), A-27, C-18, C-21, C-22, C-23, C-24, C-30, C-33, and RB-1 (analysis areas RB1, RB2, RB3, and RB4) are deterministically compliant with the Current Fire Protection Licensing Basis.
August 2011 Page S-6 SECURITY-RELATED INFORMATION - WITHHOLD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 Ameren Missouri  Callaway Plant NFPA 805 Transition Report August 2011 Page S-7 SECURITY-RELATED INFORMATION - WITHHOLD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 Table S Plant Modifications Committed Item Rank Problem Statement Proposed Modification In FPRAComp Measure Risk Informed Characterization 12-0009 L Quick response sprinkler heads are installed in cable chases A-11, C-30, and C-31. 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. Quick response sprinkler heads in cable chases A-11, C-30, and C-31 will be modified to be in accordance with the applicable requirements of NFPA 13-1976 edition. Y Y FPIP 21315 The risk from this condition is low. While the sprinkler heads do not explicitly meet NFPA code they are installed and functional and will activate in the event of a fire and provide full coverage within the fire area. Compensatory measure for NFPA 805: As required by the approved Fire Protection Program an hourly roving fire watch has been established for fire areas A-11, C-30, and C-31 which will remain in place until the sprinkler system is modified to be compliant with the NFPA code.
Compensatory measure for Current Fire Protection Licensing Basis:
As required by the approved Fire Protection Program an hourly roving fire watch has been established for fire areas A-11, C-30, and C-31 which will remain in place until the sprinkler system is modified to be compliant with the NFPA code.
FPE RAI 14 Table S-3 Implementation Items Item UnitDescriptionLAR Section / SourceAmeren MissouriCallaway Plant NFPA 805 Transition Report11-805-049Section 4.1.5.b of APA-ZZ-00741 will be revised to address that cribbing timbers 6 in. by 6 in. or larger are not required to be fire-retardant treated.4.1.2 and Attachment A 111-805-050Drawing E-2R8900 and procedure EDP-ZZ-04044 will be revised to require that, where wiring must be installed above a suspended ceiling, it shall be of a type approved in FAQ 06-0022.4.1.2 and Attachment A 111-805-051Section 4.1.3(c) of procedure APA-ZZ-00743, "Fire Team Organization and Duties," will be revised to include the requirement that industrial fire brigade members shall have no other assigned normal plant duties that would prevent immediate response to a fire or other emergency as required.4.1.2 and Attachment A 111-805-052Procedure APA-ZZ-00700 will be revised to identify that plant personnel who respond with the industrial fire brigade are trained as to their responsibilities, potential hazards to be encountered, and interfacing with the industrial fire brigade.4.1.2 and Attachment A 111-805-053OTO-ZZ-00001 and OTO-KC-00001 will be revised to incorporate credited Recovery Actions consistent with Attachment C (Fire Area Transition).4.2.1.3 and Attachment G 111-805-055Non-Power Operations risk management strategies from the NFPA 805 NSCA (Callaway Plant 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).4.2.1 and Attachment D 111-805-056Confirmation that plant modification MP 07-0151 has adequately modified the control circuitry 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.4.2.4 and Attachment C 111-805-058APA-ZZ-00700, "Fire Protection Program," will be revised to add NPO overview, definitions; road map; and risk reduction requirements for all NPO, then HRE.4.3.2 and Attachment D 111-805-059APA-ZZ-00741, "Control of Combustible Materials," will be revised to add a section which addresses outage roving fire watches with specific NPO scope.4.3.2 and Attachment D 111-805-061APA-ZZ-00703, "Fire Protection Operability Criteria and Surveillance Requirements," contains the compensatory actions to be implemented should a fire protection system required to be operable during HRE periods be found to be impaired. In these cases continuous fire watches will be implemented in the affected systems areas.4.3.2 and Attachment D 1August 2011Page S-11 LIC- 07 Table S-3 Implementation Items Item UnitDescriptionLAR Section / SourceAmeren MissouriCallaway Plant NFPA 805 Transition Report11-805-073The audit scope requirements contained in OQAM Section 18.8.e will be relocated to FSAR SP Section 9.5.1 and revised to add the Monitoring Program assessment criteria identified in Section 4.6.2. Additionally the OQAM Section 18 will be revised to change the FP QA Audit frequency from 2 years to 3 years.4.6.2 111-805-074Procedure ODP-ZZ-00002, "Equipment Status Control," Attachment 3, "Operability Evaluations," will be revised to ensure the assumed nitrogen inventory as described in Section 4.2.1.2 Safe and Stable Conditions for the Plant, is maintained in the ASD N2 accumulator tanks.4.2.1.2 111-805-075In accordance with ODP-ZZ-0016E, "Operations Technicians Watchstation Practices and Rounds," a form will be initiated to change the data points for Operations AutoTour to ensure the assumed nitrogen inventory as described in Section 4.2.1.2 Safe and Stable Conditions for the Plant, is maintained in the ASD N2 accumulator tanks.4.2.1.2 111-805-076The 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.4.1.2 and Attachment A 111-805-077FPP-ZZ-00009, "Fire Protection Training Program," will be revised to include the containment and monitoring of fire suppression agents and products of combustion in potentially contaminated areas.4.4.2 and Attachment E 111-805-078FPP-ZZ-00009, "Initial Traning Course Agenda," will be revised to include the containment and monitoring of fire suppression agents and products of combustion in potentially contaminated areas.4.4.2 and Attachment E 111-805-079FPP-ZZ-00009, "Retraining Courses and Activities," will be revised to include the containment and monitoring of fire suppression agents and products of combustion in potentially contaminated areas.4.4.2 and Attachment E 111-805-080Section 6 of HTP-ZZ-05006, "Fire Involving Radioactive Material or Entry into the Radiological Controlled Area," will be revised to address Radiation Protection actions for monitoring and control of potentially contaminated effluents.4.4.2 and Attachment E 1August 2011Page S-13 LIC- 19 Table S-3 Implementation Items Item UnitDescriptionLAR Section / SourceAmeren MissouriCallaway Plant NFPA 805 Transition Report11-805-088Configuration control mechanisms for the Fire PRA and NSCA will be revised to ensure the basis for MSO inclusion/exclusion is maintained consistent with the current plant. The rationale for excluding generically identified MSOs from the Callaway Plant Fire PRA and Callaway Plant NSCA was documented in Callaway Plant Calculation 17671-002b, "Callaway NFPA 805 Fire PRA - MSO Expert Panel Report," and Callaway Plant Calculation KC-26, "Nuclear Safety Capability Assessment," respectively. Configuration control mechanisms will be reviewed to provide reasonable confidence that the exclusion basis remains valid.4.2.1.4 and Attachment F 111-805-089The Monitoring program described in procedure EDP-ZZ-01101, "Fire Protection Monitoring Program Procedure, will be implemented after the safety evaluation issuance as part of the fire protection program transition to NFPA 805. Ameren Missouri will implement a monitoring program in accordance with FAQ 10-0059 Rev. 5 during implementation.4.6.2 111-805-090In order to adequately reflect the calculated reliability of recovery actions in the Fire HRA, the Fire HRA will be updated once procedure updates, plant modifications, and recovery action training are complete.4.2.1.3 and Attachment G 111-805-091The 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.4.1.2 and Attachment A 111-805-092An administrative control will be implemented to ensure that the breaker (PB0406) for DPAE02 is disabled open during at-power plant operation to address a potential MSO that may result in overfill of steam generators/overcooling of the RCS.4.2.1.4 and Attachment F 111-805-093The current Fire PRA does include the phase #1 version of MP 10-0038, and the AEPS, MP 10-0032. As the modification packages are completed, the FPRA will be udpated if necessary to reflect the final configuration.Attachment S 112-805-001All of the items labeled with footnote 1 in PRA RAI 1, Table 1, will be completed:F&Os: 1-7, 1-14, 1-25 , 1-3,1-13 and 1-15These commitments involve updating the fire PRA to be consistent with upgrade items that were implemented in the internal events PRA update (Revision 5) after the Callaway Plant NFPA-805 LAR was submitted.In addition, a self-assessment of the internal events PRA against the RG 1.200, Rev 2 clarifications and qualifications to determine if any gaps  exist  is in progress and will be completed, with any resolutions completed before transition to NFPA 805 occurs.Attachment U 1August 2011Page S-15 PRA RAI 01 and PRA RAI 02 MP RAI 01-a Table S-3 Implementation Items Item UnitDescriptionLAR Section / SourceAmeren MissouriCallaway Plant NFPA 805 Transition ReportAugust 2011Page S-16 FM RAI 01-f PRA RAI 03
 
Attachment T:  Revisions to Transition Report Attachment T - Clarification of
 
Prior NRC Approvals 
 
Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page T-6 Open and close control for Steam Generator D (4) Atmospheric Steam Dump Valve (ABPV0004)  Open and close control for Steam Generator D (4) AFW flow control valve from MDAFW Pump B (ALHV0005)  Open and close control for Essential Service Water to suction of TDAFP (ALHV0033)  TDAFP suction pressure indication (ALPI0026B)  Open and close control for TDAFP Governor Control valve (FCFV0313)  Open and close control for TDAFP Trip and Throttle valve (FCHV0312)  Pressurizer level indication (BBLI0460B)  Reactor Coolant System pressure indication (BBPI0406X)  Reactor Coolant System Loop 2 cold leg temperature indication (BBTI0423X)  Reactor Coolant System Loop 4 hot leg temperature indication (BBTI0443A)  Intermediate and source range neutron monitoring indication (SENI0061X and SENI0061Y) Trip and close control for Pressurizer Backup Group B breaker (PG2201)
RequestAs 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 manual actions quoted above from NUREG-0830 Supp. 3 for a Main Control Room evacuation fire event 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.SSA RAI 03 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page T-8 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 on either side of the reactor building wall which are secured by multiple pin latches. The gap between the door and the bulk heads is sealed by double-o-ring gaskets. 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 requirements consisting of a welded steel assembly with a double gasketed, flanged, and bolted cover and provided with a moveable concrete missile shield on the 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.
RequestAs 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.
SSA RAI 04 SSA RAI 04 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page T-12By 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 diesel generator day tanks are 600 gallons verse 550 gallons and that the day tank dike 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. 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 generator day tanks. The original NRC approval was granted based on the overall design of the emergency diesel generator fuel oil day tank assembly and did not solely rely on the day tank SSA RAI 05 SSA RAI 05 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page T-13dike capacity. Therefore, the basis for the prior NRC approval and the NRC conclusions made in NUREG-0830, Supplement 3, dated 05/1984 remain valid regarding acceptability of the diesel fuel oil day tank dike system in the A and B Emergency Diesel Generator rooms.
SSA RAI 05 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page T-16Prior Approval Clarification Request 6 Current Licensing Basis:Callaway Plant credits the following allowance regarding shift fire brigade staffing as stated in FSAR Standard Plant (SP) Section 16.12.1 Organization - Unit Staff  "The Unit organization shall be subject to the following: b. A site Fire Brigade of at least five members (may be less than the minimum requirements for a period of time not to exceed 2 hours in order to accommodate unexpected absence provided immediate action is taken to fill the required positions.) shall be maintained onsite at all times. The Fire Brigade shall not include the Shift Manager, and the other members of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency." Background/Basis:
The two hour grace period was originally a part of the Westinghouse Standard Technical Specifications in Section 6.2.2, Unit Staff. As stated in Section 9.5.1.6 of NRC SER NUREG 0830, "The applicant has committed to follow the staff standard technical specifications. The staff finds this acceptable." In the initial Callaway Plant Technical Specifications NPF-30, "Technical Specifications," Section 6.2.2.e contained the 2 hour grace period for fire brigade staffing. In later revisions to the Callaway Plant Technical Specifications the requirements related to fire protection were removed and relocated to the FSAR SP. The allowance for the 2 hour grace period is now located in FSAR SP Section 16.12.1 Organization - Unit Staff as stated above.
RequestAs part of this LAR submittal and transition to NFPA 805, it is requested that the NRC formally document as a "prior approval" the allowance for a 2 hour grace period for fire brigade staffing as allowed by the FSAR Section 16.12.1 statement.
LIC-16
 
Attachment V:  Revisions to Transition Report Attachment V - Fire PRA
 
Quality 
 
Table V-1 Fire PRA Peer Review - Facts and Observations (F&Os)Ameren Missouri Callaway Plant NFPA 805 Transition Report DispositionFact/Observation Status TopicF&O#Fire PRA and Internal Events PRA ClosedScientech Calculation 17671-015 provides the disposition of Internal Events F&Os.
Existing F&Os were ranked based on their fire PRA impact. Category A had the highest impact. Category B had a lesser impact. Category C had no impact.
Basis for Significance: The Fire PRA is
 
based on the internal events PRA. Changes to PRA success criteria will impact the Fire PRA.The internal events Peer Review Findings were reviewed, and their disposition documented in Attachment U to the NFPA 805 Transition LAR for Callaway.This F&O was a SUGGESTION, not a FINDING. It involves the disposition of the Internal Events PRA G AP items by the Fire PRA. The Fire PRA classified the GAP items as A, B, or C, based on the potential effect on Fire PRA results. The A and B items were resolved and incorporated into the Fire PRA while the C items were left for future updates. The C items were considered to have minimal effect on Fire PRA. The Westinghouse Peer Review Team interpreted the GAP Assessment process as a "work in progress" and suggested the work be re-reviewed and finalized. The licensee interpreted the GAP Assessment process as final at the time of t he Peer Review and did not respond to this suggestion.PRM-B2-1 August 2011 Page V-12 LIC 13 Table V-1 Fire PRA Peer Review - Facts and Observations (F&Os)Ameren Missouri Callaway Plant NFPA 805 Transition Report DispositionFact/Observation Status TopicF&O#Initiating Events - Document Applicability ClosedCallaway developed no new initiating events for the Fire PRA. T hus, the self assessment lists this as not applicable.
Basis for Significance: However, to meet this supporting requirement, a defined basis is needed to support the claim of nonapplicability of the requirements.PRM-B4-1 requires any new initiating event to be modeled in accordance with HLR-IE-A (Completeness), HLR-IE-B (Grouping) and HLR-IE-C (Frequency) from Part 2 of the ASME PRA Standard. Compliance with these HLR's is resolved as follows:HLR-IE-A (Completeness) is met because the Callaway FPRA evaluated every initiating event considered for the internal events PRA. This evaluation is shown in Table 4-1 of Report Callaway -17671-004 Fire Induced Risk Model. Further justification of completeness is not necessary.HLR-IE-B (Grouping) is not applicable because occurrence of fire events (initiating events and consequential events) are individually identified, based on the cable damage from an indivi dual fire scenario. Grouping of fire initiators (such as caused by spurious operation) is not done. HLR-IE-C (Frequency) is not applicable because frequency (or conditional probability) of fire ev ents is determined by the probabilities used for circuit analysis reflecting the cable damage in a scenario.PRM-B4-1 August 2011 Page V-13 LIC 14
 
Attachment W:  Revisions to Transition Report Attachment W - Fire PRA
 
Insights 
 
Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011 Page W-2 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 the risk increase as demonstrated in Table W-2. It is important to note that the risk reduction is 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.
PRA RAI 14 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011  Page W-3 Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF
)Scenario Description Contribution Risk Insights CCDP IF 1 CDF CLERP LERF Scenario Cumulative 1501-1A NG04C-NonVent 20.15% 20.15% Failures include MDAFP "B" 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. 1.12E-02 3.63E-04 4.07E-06 1.22E-04 4.43E-08 PRA RAI 15 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011  Page W-4 Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF
)Scenario Description Contribution Risk Insights CCDP IF 1 CDF CLERP LERF Scenario Cumulative 1501-1 NG04C-Vent 5.14% 25.29% This scenario is dominated 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. 3.72E-02 2.79E-05 1.04E-06 5.95E-04 1.66E-08 PRA RAI 15 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011  Page W-5 Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF
)Scenario Description Contribution Risk Insights CCDP IF 1 CDF CLERP LERF Scenario Cumulative YD-SXFR Startup Xfmr 4.56% 29.85% This scenario involves a large 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. 4.49E-04 2.05E-03 9.20E-07 9.93E-06 2.03E-08 PRA RAI 15 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011  Page W-6 Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF
)Scenario Description Contribution Risk Insights CCDP IF 1 CDF CLERP LERF Scenario Cumulative 4501-2B H2-Sys 4.17% 34.02% This scenario involves a large 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. 1.57E-03 5.35E-04 8.41E-07 4.12E-05 2.20E-08 PRA RAI 15 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011  Page W-7 Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF
)Scenario Description Contribution Risk Insights CCDP IF 1 CDF CLERP LERF Scenario Cumulative C10-8s NG02A-Vent 3.20% 37.22% This scenario is started by a 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. 2.45E-02 2.63E-05 6.45E-07 5.94E-04 1.56E-08 C10-17 RP140 3.15% 40.37% This scenario is started by a 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. 2.28E-02 2.79E-05 6.36E-07 5.51E-04 1.54E-08 PRA RAI 15 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011  Page W-8 Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF
)Scenario Description Contribution Risk Insights CCDP IF 1 CDF CLERP LERF Scenario Cumulative C9-12 RP139 3.09% 43.46% This scenario is started by a 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. 2.24E-02 2.79E-05 6.24E-07 5.41E-04 1.51E-08 RL015/016e RL15/16-Evac 2.25% 45.71% This scenario is a large fire in 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. 1.20E-01 3.80E-06 4.54E-07 3.37E-03 1.28E-08 PRA RAI 15 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011  Page W-9 Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF
)Scenario Description Contribution Risk Insights CCDP IF 1 CDF CLERP LERF Scenario Cumulative 3801T3 TS#3 2.21% 47.92% This scenario represents a 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 hours. Feed and bleed cooling is unavailable due to fire damage to the PORVs. 7.64E-01 5.83E-07 4.45E-07 1.09E-02 6.37E-09 4501-3 TB-Cat 2.16% 50.08% This scenario is a 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. 5.60E-02 7.79E-06 4.36E-07 1.73E-03 1.35E-08 PRA RAI 15 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011  Page W-10 Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF
)Scenario Description Contribution Risk Insights CCDP IF 1 CDF CLERP LERF Scenario Cumulative 3501T11 TS#11 1.79% 51.87% This scenario represents a 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. 1.29E-01 2.80E-06 3.61E-07 1.19E-02 3.33E-08 3801T2 TS#2 1.57% 53.44% This scenario represents a 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 hours. Feed and bleed cooling is unavailable due to fire damage to the PORV. 7.61E-01 4.17E-07 3.17E-07 1.08E-02 4.48E-09 PRA RAI 15 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011  Page W-11 Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF
)Scenario Description Contribution Risk Insights CCDP IF 1 CDF CLERP LERF Scenario Cumulative 4501-1B LO-Sys 1.48% 54.92% This scenario involves a large 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. 1.57E-03 1.90E-04 2.98E-07 4.12E-05 7.81E-09 PRA RAI 15 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011  Page W-12 Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF
)Scenario Description Contribution Risk Insights CCDP IF 1 CDF CLERP LERF Scenario Cumulative A29-WR A-29 Whole Room Burnup 1.30% 56.22% The scenario is caused by any/all fires in A-29. Fire modeling was not employed in this room. This scenario fails steam line pressure instrumentation on all steam lines causing spurious opening of SG-ASD's. Fire also fails auxiliary feedwater flow indication on several SG's. Failure of the operator to respond to the loss of instrumentation leads to loss of SG cooling and failure of feed and bleed. 2.41E-03 1.09E-04 2.63E-07 2.10E-06 2.29E-10 4203-0 TS.4203-T5 1.11% 57.33% This is the large floor-area transient fire in zone 4203. Notable fire-induced failures include the Normal Charging Pump (PBG04) via failure of bus PB03 due to electical faults. In addition, all three Normal Service Water pumps are failed. There is no fire-induced damage to safety-related equipment or offsite power. Fire risk is driven by random common cause failures of the Essential Service Water pumps to run and the electrical faults that fail the Normal Charging Pump. 1.24E-03 1.81E-04 2.24E-07 2.58E-05 4.65E-09 PRA RAI 15 Ameren Missouri Callaway Plant NFPA 805 Transition Report August 2011  Page W-13 Table W-1 Callaway Significant Fire Initiating Events (Representing all Fire Scenarios Contributing More than 1% of CDF
)Scenario Description Contribution Risk Insights CCDP IF 1 CDF CLERP LERF Scenario Cumulative A13-WR A-13 Whole Room Burnup 1.01% 58.34% This scenario is caused by any/all fires in A-13; fire modeling was not employed in this area. Fire-induced damage includes potentially spuriously open Steam Generator "A", "B", and "D" atmospheric steam dump valves ("B" is not recoverable from the MCR), failure of the Motor-Driven AFW Pump "B" and the Turbine-Driven AFW Pump, and potentially spuriously open Steam Generator Blowdown valves "B" and "C". Due to relatively significant damage to the AFW system, risk is dominated by failure of AFW and subsequent failure of Feed and Bleed. 8.77E-04 2.33E-04 2.04E-07 7.67E-07 1.78E-10 Note1IgnitionFrequency(IF)includesseverityfactorandprobabilityofnonsuppression,whereapplicable PRA RAI 15 Attachment W - Table W-2 Fire Area Risk Summary Fire AreaArea DescriptionNFPA 805 BasisFire Area CDF/LERF VFDR (Yes/No)Ameren MissouriCallaway Plant NFPA 805 Transition Report RAs (Yes/No)Fire Risk EvalCDF/LERFA-288.27E-091.97E-10 /8.27E-091.97E-10Auxiliary Shutdown Panel Section A4.2.4.2 Yes No / A-292.63E-072.29E-10 /1.28E-081.12E-11Auxiliary Feedwater Valve Compartment, SG A&D4.2.4.2 Yes Yes / A-308.81E-087.75E-11 /0.00E+000.00E+00Auxiliary Feedwater Valve Compartment, SG B&C4.2.4.2 Yes Yes / A-331.50E-081.32E-11 /3.86E-093.39E-12Auxiliary Shutdown Panel Section B4.2.4.2 Yes No / AB-1 /Auxiliary Boiler Room4.2.3.2 No NoN/AC-16.96E-126.84E-14 /6.96E-126.84E-14Pipe Space and Tank Area4.2.4.2 Yes No / C-2 /Control Building North Cable Chase4.2.3.2 No NoN/AC-3 /Control Building Cable Chase B4.2.3.2 No NoN/AC-52.29E-102.38E-12 /Control Building Access Control Area4.2.3.2 No NoN/AC-63.51E-103.86E-12 /Control Building Access Control Area4.2.3.2 No NoN/AC-72.69E-092.19E-11 /2.69E-092.19E-11Control Building North Cable Chase4.2.4.2 Yes No / C-81.15E-111.65E-14 /Control Building Cable Chase B4.2.3.2 No NoN/AC-99.31E-072.10E-08 /1.09E-072.45E-09Switchgear Room A4.2.4.2 Yes No / August 2011Page W-16 6.96 E-1 2 6.8 4E-14 /Y es4.2.4.2 PRA RAI 04-f Attachment W - Table W-2 Fire Area Risk Summary Fire AreaArea DescriptionNFPA 805 BasisFire Area CDF/LERF VFDR (Yes/No)Ameren MissouriCallaway Plant NFPA 805 Transition Report RAs (Yes/No)Fire Risk EvalCDF/LERFD-21.10E-087.23E-11 /Diesel Generator B4.2.3.2 No NoN/A FB-15.39E-081.99E-11 /4.56E-087.98E-12Fuel Handling Building4.2.4.2 Yes No / LDF-13.78E-095.42E-12 /Laundry Decontamination Facility4.2.3.2 No NoN/ARB-12.36E-071.93E-09 /2.30E-086.23E-10Reactor Building General Area4.2.4.2 Yes Yes / RSB-1 /RAM Storage Building4.2.3.2 No NoN/ARW-13.76E-085.40E-11 /Radwaste Building4.2.3.2 No NoN/ATB-16.54E-061.36E-07 /0.00E+000.00E+00Turbine Building General Area4.2.4.2 Yes No / UNCT2.73E-094.61E-13 /Ultimate Heat Sink Cooling Tower A, El. 2000'4.2.3.2 No NoN/A UNPH2.69E-092.62E-11 /Essential Service Water Pump Room A4.2.3.2 No NoN/A USCT2.73E-094.61E-13 /Ultimate Heat Sink Cooling Tower B, El. 2000'4.2.3.2 No NoN/A USPH2.72E-092.69E-11 /Essential Service Water Pump Room B4.2.3.2 No NoN/AYD-11.03E-062.18E-08 /1.68E-082.94E-12Plant Yard Area El. 2000'4.2.4.2 Yes No / 2.03E-053.99E-07 /Total1.96E-064.11E-08 / August 2011Page W-191.96E-0 6 4.11E-0 8 /PRA RAI 04-f
 
Attachment X:  Revisions to Transition Report Attachment X - Other Requests
 
for Approval
 
Approval is requested for the 20 foot separation zones in Fire Areas A-1, A-16, A-27 and C-1 as complying with the criteria of "no intervening combustible materials or fire hazards" and meeting the deterministic compliance criteria of NFPA 805 section 4.2.3.3(b).          Callaway Plant Fire Areas A-1, A-16, A-27, and C-1, credit NFPA Section 4.2.3.3(b) to achieve deterministic compliance. NFPA 805 Section 4.2.3.3(b) requires the following:
 
As noted in NRC Generic Letter (GL) 86-10, 20 feet of separation with absolutely no intervening combustible or fire hazards is not always achievable. Therefore, engineering judgment is necessary to evaluate each specific configuration. GL 86-10 provided the general guidance regarding evaluation of intervening combustibles and fire hazards within a 20 ft separation zone. This evaluation will consider in-situ combustibles and ignition sources to determine if they represent intervening combustibles or a fire hazard within the 20 ft zone. As noted in GL 86-10, transient combustibles are not considered as an intervening combustible and they are addressed by the Callaway transient combustible control program. All of the 20 ft separation zones evaluated herein are in the current pre-transition Fire Protection Program, and the 20 ft separation zones are clearly marked on the floor of the respective fire areas with red striping and "No Combustible Zone" lettering. The raceway cables in the 20 ft separation zone raceways are all IEEE-383 qualified Thermoset type cable. Per the GL 86-10 guidance, cables in conduit and cables in completely enclosed cable trays are not considered intervening combustibles. At Callaway, enclosed cable trays are cable trays with both top and bottom tray covers and the tray cover seams are sealed with a standard fire resistive caulk such as 3M CP
 
25 (Ref E-2R8900 Sht. 53). 
 
Additionally, small items like equipment tags, signage, labels, fire extinguisher tags, hose rack covers, postings comprised of combustible materials (plastic / paper), gai-tronics handsets, and emergency light units exist in small quantities in the areas. However, they are not considered to exist in sufficient quantity to propagate or spread fire across a 20 ft separation zone therefore these are not considered intervening combustibles or fire hazards.   
 
Fire Area C-1 To meet deterministic separation criteria, Fire Area C-1 Fire Zone 3101 is divided into two safe shutdown analysis areas C-1N and C-1S which are separated by a 20 ft separation zone. As shown in Table 4-3 this zone has automatic detection and automatic suppression and the LIC-18 location of the 20 ft separation zone is shown on plant drawing A-2817. Additionally, the redundant circuits are separated by greater than 20 feet. Fire Area C-1 is a piping chase with primarily piping, valves and conduit. There are no ignition sources as defined by NUREG/CR-6850 located within the 20 ft zone and there are no cable trays within the 20 ft separation zone. There are emergency light units and several small (<5hp motor) MOVs in the 20 ft separation zone. Based on the plant configuration, the 20 ft separation zone has no intervening combustibles or fire hazards and the 20 ft separation zone in fire area C-1 is considered to meet the criteria of NFPA 805 Section 4.2.3.3(b). Fire Area A-1 To meet deterministic separation criteria, Fire Zones 1206 and 1207 have a 20-foot separation zone. As shown in Table 4-3 this zone has automatic detection and automatic suppression and the location of the 20 ft separation zone is shown on drawing A-2818. Additionally, the redundant circuits are separated by greater than 20 feet. These zones consist of a pipe chase area on the 1988 elevation of the Auxiliary Building containing primarily piping, valves and conduit. There are no ignition sources as defined by NUREG/CR-6850 located within the 20 ft zone and there are no cable trays within the 20 ft separation zone. There are several small AOVs in the 20 ft separation zone. Based on the plant configuration, the 20 ft separation zone has no intervening combustibles or fire hazards and the 20 ft separation zone in fire area A-1 is considered to meet the criteria of NFPA 805 Section 4.2.3.3(b). 
 
Fire Area A-16 To meet deterministic separation criteria, Fire Area A-16, Fire Zone 1408 is divided into two safe shutdown analysis areas A-16N and A-16S which are separated by a 20 foot separation zone.
As shown in Table 4-3 this zone has partial automatic detection and automatic suppression over portions of the 20 ft separation zone. Additionally, the redundant circuits and equipment are separated by much greater than 20 feet. This fire zone is the general corridor area of the Auxiliary Building 2026 elevation and the separation zone is established to separate the two trains of Component Cooling Water components that are located in the large fire area. The location of the 20 ft separation zone is shown on drawing A-2814. During initial Callaway Plant licensing, the configuration of the 20 ft separation zone was reviewed and approved by the NRC in Callaway Plant SSER 3 and that approval is being carried forward (Ref. Attachment K Licensing Action 008). The approval in SSER 3 was stated as follows; 
 
SSER 3 states:    The cable tray fire stops consist of enclosed cable trays with a section of RTV foam inside the cable tray as shown on plant drawing E-2R8900-Sheet 65. All cable trays that traverse through LIC-18 the 20 ft separation zone are fully enclosed. There are no ignition sources as defined by NUREG/CR-6850 located within the 20 ft separation zone. Only minor combustibles such as a hose station with a cover, a bank of Halon System Bottles, and a small (<5hp) MOV are in the 20 ft separation zone. The 20 ft separation zone has no intervening combustibles or fire hazards and the 20 ft separation zone in Fire Area A-16 is considered to meet the criteria of NFPA 805 Section 4.2.3.3(b) based on prior approval. 
 
Fire Area A-27 To meet deterministic separation criteria, Fire Zone 1403 has a 20 ft separation zone. As shown in Table 4-3 this zone has automatic detection and automatic suppression and the location of the 20 foot separation zone is shown on drawing A-2814. Additionally, the redundant circuits and equipment are separated by much greater than 20 feet. Fire Zone 1403 contains the Control Rod Drive MG Sets and the Reactor Trip Switchgear along with numerous electrical panels and their associated cable trays. As shown in Figure 1, there are cable trays, a motor control center, an air handling unit, and one of the control rod drive motor generator sets located within the 20 ft separation zone.   
 
The two raceways containing redundant trains of cables are separated by approximately 58 ft.
Figure 1 shows the location of the two redundant raceways relative to the 20 ft separation zone and to each other. One set of redundant raceways are cable trays that enter the fire area from the south wall, travel approximately 8 ft where the cable trays turn west to exit through a fire rated penetration seal into the Control Building. The bottom trays are approximately 15 ft above the floor. The second redundant raceway is a conduit that crosses the zone at height 18 ft
 
above the floor. 
 
A 480 volt non-safety related motor control center (MCC) PG20G is located in the 20 ft separation zone. As shown in Figure 1, the cables in tray exiting this MCC 1) route to the cable
 
trays that do not cross through the 20 ft separation zone or 2) route to cable trays which are fully enclosed, so they pose no concern as an intervening combustible. Based on the configuration the MCC is not considered a fire hazard that can result in a fire that will propagate across the 20 ft separation zone. 
 
Air handling unit SGL20, the air handling unit for the room itself, is located within the 20 ft separation zone. The air handling unit is not located near secondary combustibles. Based on
 
the distance from the cable trays and the insignificant amount of combustibles associated with the unit, the air handling unit is not considered a fire hazard that can result in a fire that will propagate across the 20 ft separation zone. 
 
Control Rod Drive Motor Generator set SF01 is located within the 20 ft separation zone but located greater than 17 feet from the nearest redundant train raceway. The motor generator set has exposed cables in tray that route to MCC SF103A which is also located in the 20 ft separation zone. Figure 1 shows the location of these two components. There are no exposed secondary combustibles near the motor generator set or its MCC that if ignited can propagate fire across the 20 ft separation zone. Based on the configuration the MCC and SF01 are not intervening combustibles or a fire hazard that can result in a fire that will propagate across the 20 ft separation zone. 
 
As shown in Figure 1 there are two sets of non-safety related cable trays that traverse the 20 ft separation zone. As previously discussed, all the exposed cables are Thermoset type IEEE-383 qualified cables. The set of cable trays on the west side of the 20 ft separation zone are located approximately 2'-6" from the west wall and contain three horizontally stacked cable LIC-18 trays. These cable trays are fully enclosed for over 50 ft of the tray length to a point beyond the redundant conduit raceway location. The enclosed portion of the cable trays extends into the 20 ft separation zone approximately 8 ft. The set of three cable trays on the east side of the 20 ft separation zone are fully enclosed for greater than 50 ft of the tray length to a point beyond the redundant conduit raceway location. The enclosed portion of the cable trays extend into the 20 ft separation zone for approximately 4 feet and the ca ble trays turn and exit the fire area and do not cross through the 20 ft separation zone. 
 
In the A-27 fire area, the 20 ft separation zone has fixed ignition sources and combustibles in the form of exposed cable. However, the separation of required cables and equipment of redundant success paths is by a horizontal distance of approximately 58 ft and automatic fire detectors and an automatic fire suppression system is installed throughout the fire area.
Horizontal stacks of cable trays are fully enclosed for significant portions of their travel path in
 
the 20 ft separation zones thereby eliminating the possibility a fire can propagate across the 20 ft separation zone via the cable trays. The fixed ignition sources that exist in the 20 ft separation zone are located such that their ignition will not result in a fire that can propagate
 
across the 20 ft separation zone. There are no intervening combustibles or fire hazards in the 20 ft separation zone therefore, the 20 ft separation zone in Fire Area A-27 is considered to
 
meet the criteria of NFPA 805 Section 4.2.3.3(b). 
 
LIC-18 Based on the analysis presented above, Fire Areas A-1, A-16, A-27 and C-1 all comply with the criterion of NFPA 805 Section 4.2.3.3. The evaluation shows that there is adequate separation along with automatic detection and suppression to ensure that one train of the redundant trains of circuits that exist within each of the fire areas remains free of fire damage. The design meets deterministic criteria therefore this is not a performance based approval request.
 
Based on the analysis presented above, Fire Areas A-1, A-16, A-27 and C-1 all comply with the criterion of NFPA 805 Section 4.2.3.3. The evaluation shows that there is adequate separation along with automatic detection and suppression to ensure that one train of the redundant trains of circuits that exist within each of the fire areas remains free of fire damage. The design meets deterministic criteria therefore this is not a performance based approval request.
 
NRC approval is requested to document that the 20 ft separation zones within Fire Areas A-1, A-16, A-27 and C-1 as described and evaluated above comply with the criteria of "no
 
intervening combustible materials or fire hazards" and meet the deterministic compliance criteria of NFPA 805 Section 4.2.3.3(b). LIC-18  to ULNRC-05876
 
Page 1 of 1 Enclosure 2, Computer Input Files for Fire Modeling RAIs 2b, 3b and 3t  The following files are provided on CD:
* FDS computer input file reque sted for Fire Modeling RAI 2b
* CFAST computer input files reque sted for Fire Modeling RAI 3b
* FDS computer input files requested for Fire Modeling RAI 3t}}

Revision as of 03:16, 18 September 2018

Callaway, Unit 1 - Response to Request for Additional Information Re Adoption of National Fire Protection Association Standard NFPA 805
ML12194A624
Person / Time
Site: Callaway Ameren icon.png
Issue date: 07/12/2012
From: Maglio S
Ameren Missouri, Union Electric Co
To:
Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML121940741 List:
References
ULNRC-05876
Download: ML12194A624 (4)
v • d • e


Text

WAmeren MISSOURI July 12, 2012 ULNRC-05876 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Ladies and Gentlemen:

10 CFR 50.90 DOCKET NUMBER 50-483 CALLA WAY PLANT UNIT 1 UNION ELECTRIC CO. FACILITY OPERATING LICENSE NPF -30 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING ADOPTION OF NATIONAL FIRE PROTECTION ASSOCIATION STANDARD NFPA 805 Callaway Plant Reference 1 ): Ameren Missouri Letter ULNRC-05851, Response to Request for Additional Information Regarding Adoption ofNational Fire Protection Association Standard NFP A 805 dated April 17, 2012 (T AC No. ME7046) On March 2, 2012, Callaway Plant formally received from the NRC Staff a request for additional information (RAI) related to the application request to adopt a new fire protection licensing basis based on NFPA 805 which complies with the requirements in 10 CFR 50.48(a) and (c) and the guidance in Revision 1 of Regulatory Guide (RG) 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants". On April17, 2012 Ameren Missouri transmitted to the NRC Staff via Reference 1, responses to a portion of the RAI questions.

On April 12th, 17th, & June 21th, 2012, teleconferences between Ameren Missouri and the NRC staff reviewers were conducted in which additional requests/questions on the remaining RAI's were discussed.

Following these teleconferences the NRC provided to Ameren Missouri supplemental requests/questions to the original RAI's. The purpose of this letter is to provide the additional responses to the RAI transmittal dated March 2, 2012 including responses to the supplemental requests/questions identified during the teleconferences.

PO Box 620 Fulton, MD 65251 AmerenMissouri.com ULNRC-05876 July 12, 2012 Page2 Enclosure 1 provides the RAI responses not provided in Reference

1. Enclosure 1 also contains a description of changes to the license amendment request (LAR) that Ameren Missouri has identified as being warranted.

Attachments are provided within the enclosure to show the LAR Transition Report changes from the original submittal.

Enclosure 2 provides the computer program input files requested in Fire Modeling RAis 2b, 3b and 3t. This submittal does contain revised/new commitments which are provided in the revised LAR Transition Report Attachment S, Plant Modifications and Items to be Completed during Implementation.

The No Significant Hazards Consideration determination provided in the original submittal is not altered by the additional information provided in this letter. If there are any questions, please contact Mr. Scott Maglio at (573) 676-8719 or Mr. Roger Wink at (314) 225-1561.

I declare under penalty of peijury that the foregoing is true and correct. Executed on: '1 ftL-I U 11..-.RCW/nls Sincerely, Regulatory Affairs Manager Enclosure 1, Request for Additional Information (RAI) with Callaway Plant Response Enclosure 2, Computer input files for Fire Modeling RAis 2b, 3b and 3t.

ULNRC-05876 July 12, 2012 Page 3 cc: U.S. Nuclear Regulatory Commission (Original and 1 copy) Attn: Document Control Desk Washington, DC 20555-0001 Mr. Elmo E. Collins Regional Administrator U.S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. Andrew Hon Senior Project Manager, Callaway Plant Office ofNuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-8G 14 Washington, DC 20555-2738 ULNRC-05876 July 12, 2012 Page4 Index and send hardcopy to QA File A160.0761 Hardcopy:

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

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

A. C. Heflin F. M. Diya C. 0. Reasoner III D. W. Neterer L. H. Graessle J. S. Geyer S. A. Maglio R. Holmes-Bobo NSRB Secretary T. B. Elwood Mr. Bill Muilenburg (WCNOC) Mr. Tim Hope (Luminant Power) Mr. Ron Barnes (APS) Mr. Tom Baldwin (PG&E) Mr. Mike Murray (STPNOC) Ms. Linda Conklin (SCE) Mr. John O'Neill (Pillsbury Winthrop Shaw Pittman LLP) Missouri Public Service Commission Mr. Dru Buntin (DNR)

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