ML13210A449

Issuance of Amendment Regarding Transition to a Risk-Informed Performance-Based Fire Protection Program in Accordance with 10 CFR 50.48(c)

ML13210A449
Person / Time
Site:	Duane Arnold
Issue date:	09/10/2013
From:	Feintuch K Plant Licensing Branch III
To:	Richard Anderson NextEra Energy Duane Arnold
Faria C
References
TAC ME6818
Download: ML13210A449 (159)

Text

,.,

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 10, 2013 Mr. Richard L. Anderson Vice President NextEra Energy Duane Arnold Energy Center 3277 DAEC Road Palo, IA 52324-9785

SUBJECT:

DUANE ARNOLD ENERGY CENTER - ISSUANCE OF AMENDMENT REGARDING TRANSITION TO A RISK-INFORMED, PERFORMANCE-BASED FIRE PROTECTION PROGRAM IN ACCORDANCE WITH 10 CFR 50.48(c)

(TAC NO. ME6818)

Dear Mr. Anderson:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 286 to Renewed Facility Operating License No. DPR-49 for the Duane Arnold Energy Center (DAEC). The amendments consist of changes to the licenses and Technical Specifications (TSs) in response to your application dated August 5,2011, as supplemented by letters dated October 14, 2011, April 23, 2012, May 23, 2012, July 9, 2012, October 15, 2012, January 11, 2013, February 12, 2013, March 6,2013, May 1, 2013, May 29, 2013, two supplements dated July 2,2013, August 5,2013, and August 28,2013. NextEra Energy Duane Arnold, LLC (NextEra or the licensee), submitted a license amendment request (LAR) to allow the licensee to maintain a fire protection program in accordance with Title 10 of the Code of Federal Regulations (10 CFR), Section 50.48(c) for DAEC and change the license and TSs accordingly.

The proposed amendment would transition the DAEC fire protection program to a risk-informed, performance-based program based on National Fire Protection Association (NFPA) 805, in accordance with 10 CFR 50.48(c). NFPA 805 allows the use of performance-based methods

. such as fire modeling and risk-informed methods such as fire probabilistic risk assessment to demonstrate compliance with the nuclear safety performance criteria.

The fire protection license condition in DAEC's is Renewed Facility Operating License No.

DPR-49 Condition is revised to reflec~ use of NFPA 805. To assure proper pagination of the license, the NRC is issuing license pages 3 through 7, but the only changes are the changes to the fire protection license condition.

R. Anderson

- 2 A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice Sincerely,

-~~~~

Karl D. Fe~, Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-331

Enclosures:

1. Amendment No. 286 to DPR-49

2. Safety Evaluation cc w/encls: Distribution via ListServ

ENCLOSURE 1 AMENDMENT NO. 286 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-49 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER DOCKET NO. 50-331

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NEXTERA ENERGY DUANE ARNOLD, LLC DOCKET NO. 50-331 DUANE ARNOLD ENERGY CENTER AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 286 License No. DPR-49

1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by NextEra Energy Duane Arnold, LLC (the licensee) dated August 5, 2011, as supplemented by letters dated October 14, 2011, April 23, 2012, May 23,2012, July 9,2012, October 15, 2012, January 11,2013, February 12,2013, March 6,2013, May 1,2013, May 29, 2013, two supplements dated July 2,2013, August 5,2013, and August 28, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter'l; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

- 2

2.

Accordingly, the license is hereby amended as indicated in the attachment to this license amendment, and paragraph 2.C(3) of Renewed Facility Operating License No. DPR-49 Condition 2.C(3) is amended to read as follows:

(3)

Fire Protection Program NextEra Energy Duane Arnold, LLC shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated August 5, 2011 (and supplements dated October 14, 2011, April 23, 2012, May 23,2012, July 9,2012, October 15, 2012, January 11, 2013, February 12, 2013, March 6, 2013, May 1, 2013, May 29, 2013, two supplements dated July 2, 2013, August 5, 2013, and August 28, 2013), and as approved in the safety evaluation report dated September 10, 2013. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior !\\IRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed Fire PRA (FPRA) model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic' methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.

(a) Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense-in-depth philosophy and must.maintain.

sufficient safety margins. The change may be implemented following completion of the plant change evaluation.

(b) Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1 x1 0-7/ year (yr) for CDF and less than 1x10-B/yr for LERF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain

- 3 sufficient safety margins. The change may be implemented following completion of the plant change evaluation.

Other Changes that May Be Made Without Prior NRC Approval

1. Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program and Design Elements. Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3, element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard.

The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3, elements are acceptable because the alternative is "adequate for the hazard." Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are as follows:

• "Fire Alarm and Detection Systems" (Section 3.8);

• "Automatic and Manual Water-Based Fire Suppression Systems" (Section 3.9);

• "Gaseous Fire Suppression Systems" (Section 3.10); and,

• "Passive Fire Protection Features" (Section 3.11).

This License Condition does not apply to any demonstration of equivalency under Section 1.7 of NFPA 805.

2. Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in the NRC safety evaluation report dated September 10, 2013, to

-4 determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.

Transition License Conditions (1) Before achieving full compliance with 10 CFR 50.48( c), as specified by (2) and (3) below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (2) above.

(2) The licensee shall implement the modifications to its facility, as described in, Attachment S, Table S-1, "Plant Modifications Committed," of DAEC letter NG-13-0287, dated July 2, 2013, to complete the transition to full compliance with 10 CFR 50.48(c) by December 31,2014. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

(3) The licensee shall implement the items listed in Enclosure 2, Attachment S, Table S-2, "Implementation Items," of DAEC letter NG-13-0287, dated July 2, 2013, by March 9, 2014.

3.

This license amendment is effective as of its date of issuance and shall be implemented by March 9, 2014.

FOR THE NUCLEAR REGULATORY COMMISSION Robert D. Carlson, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. DPR-49 Date of Issuance: September 10, 2013

ATTACHMENT TO LICENSE AMENDMENT NO. 286 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 Replace the following page of Renewed Facility Operating License DPR-49 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

REMOVE INSERT 3 through 5 3 through 7 Replace the following pages of Appendix A, Technical Specifications, with the attached revised pages. 'The revised pages are identified by.amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT 5.0-6 5.0-6

- 3 C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 _CFR Chapter I; Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level NextEra Energy Duane Arnold, LLC is authorized to operate the Duane Arnold Energy Center at steady state reactor core power levels not in excess of 1912 megawatts (thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 286, are hereby incorporated in the license. NextEra Energy Duane Arnold, LLC shall operate the facility in accordance with the Technical Specifications.

(a) For Surveillance Requirements (SRs) whose acceptance criteria are modified,

. either directly or indirectly, by the increase in authorized maximum power level in 2.C.(1) above, in accordance with Amendment No. 243 to Facility Operating License DPR-49, those SRs are not required to be performed until their next scheduled performance, which is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment No. 243.

(b) Deleted.

(3) Fire Protection Program NextEra Energy Duane Arnold, LLC shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated August 5,2011 (and supplements dated October 14, 2011, April 23, 2012, May 23,2012, July 9,2012, October 15, 2012, January 11,2013, February 12, 2013, March 6, 2013, May 1, 2013, May 29,2013, two supplements dated July 2,2013, and supplements dated August 5, 2013 and August 28, 2013) and as approved in the safety evaluation report dated September 10, 2013. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

Renewed License No. DPR-49 Amendment 286

- 4 Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.

(a) Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense in-depth philosophy and must maintain sufFicient safety margins. The change may be implemented following completion of the plant change evaluation.

(b) Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1 x10-7/year (yr) for CDF and less than 1 x10-8/yr for LERF: The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.

Other Changes that May Be Made Without Prior NRC Approval

1. Cha'nges to NFPA 805, Chapter 3, Fundamental Fire Protection Program. Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to NFPA 805, Chapter 3 element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard.

The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3 elements are acceptable because the alternative is "adequate for the hazard." Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3; are as follows:

Renewed License No. DPR-49 Amendment 286

- 5 Fire Alarm and Detection Systems (Section 3.8);

Automatic and Manual Water-Based Fire Suppression Systems (Section 3.9);

Gaseous Fire Suppression Systems (Section 3.10);,and, Passive Fire Protection Features (Section 3.11 ).

This License Condition does not apply to any demonstration of equivalency under Section 1.7 of NFPA 805.

2. Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in the NRC safety evaluation report dated September 10, 2013 to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.

Transition License Conditions (1) Before achieving full compliance with 10 CFR 50.48(c), as specified by (2) and (3) below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (2) above.

(2) The licensee shall implement the modifications to its facility, as described in, Attachment S, Table S-1, "Plant modifications Committed," of DAEC letter NG-13-0287, dated July 2,2013, to complete the transition to full compliance with 10 CFR 50.48(c) by December 31,2014. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

(3) The licensee shall implement the items listed in Enclosure 2, Attachment S, Table S-2, "Implementation Items," of DAEC letter NG-13-0287, dated July 2, 2013, by March 9, 2014.

(4) The licensee is authorized to operate the Duane Arnold Energy Center following installation of modified safe-ends on the eight primary recirculation system inlet lines which are described in the licensee letter dated July 31, 1978, and supplemented by letter dated December 8, 1978.

(5) Physical Protection NextEra Energy Duane Arnold, LLC shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, Renewed License No. DPR-49 Amendment 286

- 6 and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.SS (S1 FR 27817 and 27822) and to the authority of 10 CFR SO.90 and 10 CFR SO.S4(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Duane Arnold Energy Center Physical Security Plan," submitted by letter dated May 16,2006.

NextEra Energy Duane Arnold, LLC shall fully implement and maintain in effect all provisions of the Commission-approved Duane Arnold Energy Center/NextEra Energy Duane Arnold, LLC Cyber Security Plan (CSP), including changes made pursuant to the authority of 10 CFR SO.90 and 10 CFR SO.S4(p). The Duane Arnold Energy Center/NextEra Energy Duane Arnold, LLC CSP was approved by License Amendment No. 278. as supplemented by a change approved by license Amendment No. 284.

(6) Deleted (7) Additional Conditions The Additional Conditions contained in Appendix B. as revised through Amendment No. 279, are hereby incorporated into this license. NextEra Energy Duane Arnold.

LLC shall operate the facility in accordance with the Additional Conditions.

(8) The licensee is authorized to revise the Updated Final Safety Analysis Report by deleting the footnote for Section 9.1.4.4.5 which states: u*The NRC has not endorsed the reactor building crane as single-failure proof (Reference 9)," and by deleting Reference 9 of the references for Section 9.1.

(9) Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a) Fire fighting response strategy with the following elements:

1.

Pre-defined coordinated fire response strategy and guidance

2.

Assessment of mutual aid fire fighting assets

3.

Designated staging areas for equipment and materials

4.

Command and control

5.

Training of response personnel (b) Operations to mitigate fuel damage considering the following:

1.

Protection and use of personnel assets

2.

Communications

3.

Minimizing fire spread

4.

Procedures for implementing integrated fire response strategy

5.

Identification of readily-available pre-staged equipment

6.

Training on integrated fire response strategy

7.

Spent fuel pool mitigation measures (c) Actions to minimize release to include conSideration of:

1.

Water spray scrubbing

2.

Dose to onsite responders (10) The licensee shall implement and maintain all Actions required by Attachment 2 to NRC Order EA-06-137, issued June 20. 2006, except the last action that requires incorporation of the strategies into the site security plan, contingency plan, emergency plan and/or guard training and qualification plan, as appropriate.

Renewed License No. DPR-49 Amendment 286

- 7 (11) The information in the UFSAR supplement, as revised, submitted pursuant to 10 CFR 54.21 (d), shall be incorporated into the UFSAR no later than the next scheduled update required by 10 CFR 50.71(e) following the issuance of this renewed operating license. Until this update is complete, the licensee may not make changes to the information in thesLipplement. Following incorporation into the UFSAR, the need for prior Commission approval of any changes will be governed by 10 CFR 50.59.

(12) The UFSAR supplement, as revised, submitted pursuant to 10 CFR 54.21(d),

describes certain future activities to be completed prior to and/or during the period of extended operation. The licensee shall complete these activities in accordance with Appendix A of NUREG-1955, "Safety Evaluation Report Related to the License Renewal of Duane Arnold Energy Center," dated November 2010, as supplemented by letter from the licensee to the NRC dated November 23, 2010. The licensee shall notify the NRC in writing when activities to be completed prior to the period of extended operation are complete and can be verified by NRC inspection.

(13) The licensee shall implement the most recent staff-approved version of the Boiling Water Reactor Vessels and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) as the method to demonstrate compliance with the requirements of 10 CFR Part 50, Appendix H. Any changes to the BWRVIP ISP capsule withdrawal schedule must be submitted for staff review and approval. Any changes to the BWRVIP ISP capsule withdrawal schedule which affects the time of withdrawal of any surveillance capsules must be incorporated into the licensing 'basis. If any surveillance capsules are removed without the intent to test them, these capsules must be stored in a manner which maintains them in a condition which would support re-insertion into the reactor pressure vessel if necessary.

D. This license is effective as of thedate of issuance and shall expire at midnight February 21, 2034.

FOR THE NUCLEAR REGULATORY COMMISSION Original signed by Eric J. Leeds Eric J. Leeds, Director Office of Nuclear Reactor Regulation

Enclosures:

1. Appendix A Technical Specifications

2. Appendix B Additional Conditions Date of Issuance: December 16, 2010 Renewed License No. DPR-49 Amendment 286

5.4 Procedures 5.0 ADMINISTRATIVE CONTROLS 5.4 Procedures 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

a.

The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A. February 1978;

b.

The emergency operating procedures required to implement the requirements of NUREG-0737 and to NUREG-0737. Supplement 1, as stated in Generic Letter 82-33;

c.

Quality assurance for effluent and environmental monitoring;

d.

[Deleted]; and

e.

All programs specified in Specifications 5.5.

DAEC 5.0-6 Amendment 286

ENCLOSURE 2

~.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION TRANSITION TO A RISK-INFORMED, PERFORMANCE-BASED FIRE PROTECTION PROGRAM IN ACCORDANCE WITH 10 CFR 50.48(c)

AMENDMENT NO. 286 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-49 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER DOCKET NO. 50-331

TABLE OF CONTENTS SAFETY EVALUATION

1.0 INTRODUCTION

- 1 1.1 Background...........................................................................................................

- 1 1.2 Requested Licensing Action..................................................................................

- 3

2.0 REGULATORY EVALUATION

- 4 2.1 Applicable Regulations..........................................................................................

- 7 2.2 Applicable Staff Guidance.....................................................................................

- 8 2.3 NFPA 805 Frequently Asked Questions................................................................

- 12 2.4 Orders, License Conditions, and Technical Specifications....................................

- 14 2.4.1 Orders................................................................................................................ :...

-14 2.4.2 License Conditions................................................................................................

- 15 2.4.3 Technical Specifications........................................................................................

- 16 2.4.4 Updated Final Safety Analysis Report (UFSAR)....................................................

- 16 2.5 Rescission of Exemptions......................................................................................

- 16 2.6 Self-Approval Process for FPP Changes (Post-Transition)...................................

- 18 2.6.1 Post-Implementation Plant Change Evaluation Process........................................

- 19 2.6.2 Requirements for the Self Approval Process Regarding Plant Changes................

- 21 2.7 Implementation................................. :....................................................................

- 24 2.7.1 Modifications.........................................................................................................

- 24 2.7.2 Schedule................................................................................................................

- 24 2.8 Summary of Implementation Items.......................................................................

- 25

3.0 TECHNICAL EVALUATION

......................... *......................................................... - 25 3.1 NFPA 805 Fundamental FPP and Design Elements.............................................

- 26 3.1.1 Compliance with NFPA 805, Chapter 3 Requirements.. :............................... :.......

- 27 3.1.1.1 Compliance Strategy - Complies............................................................

- 28 3.1.1.2 Compliance Strategy - Complies with Clarification...................................

- 30 3.1.1.3 Compliance Strategy - Complies with Use of EEEEs..............................

- 32 3.1.1.4 Compliance Strategy - Complies with Previous !\\IRC Approval...............

- 32 3.1.1.5 Compliance Strategy - Submit for NRC Approval...................................

- 32 3.1.1.6 Compliance Strategy - Multiple Strategies...............................................

- 33

- ii 3.1.1.7 Chapter 3 Sections Not Reviewed...........................................................

- 33 3.1.1.8 Compliance with Chapter 3 Requirements Conclusion............................

- 34 3.1.2 Identification of the Power Block............................................................................

- 34 3.1.3 Closure of Generic Letter 2006-03, "Potentially Nonconforming Hemyc' and MpM Fire Barrier Configurations," Issues........................................................ :.....

- 35 3.1.4 Performance Based Methods for NFPA 805, Chapter 3, Elements....... ;................

- 35 3.1.4.1 NFPA 805, Section 3.2.3(1) - Inspection, Testing, and Maintenance Proced ures.............................................................................................. - 35 3.1.4.2 NFPA 805, Section 3.3.3 - Interior Finish...............................................

- 37 3.1.4.3 NFPA 805, Section 3.3.5.2 - Metal Tray and Metal Conduit for Electrical Raceways..................... :........................ :.................................

- 39 3.1.4.4 NFPA 805, Section 3.5.11 - Common Isolation of Fire Water Supply to Fixed Systems and Backup Manual Hose Stations............. :....................

- 40 3.2 Nuclear Safety Capability Assessment (NSCA) Methods......................................

- 42 3.2.1 Compliance with NFPA 805 NSCA Methods..........................................................

- 44 i

'I 3.2.1.1 Attribute Alignment - Aligns.....................................................................

- 46 3.2.1.2 Attribute Alignment - Aligns with Intent...................................................

- 46 3.2.1.3 Attribute Alignment - Not in Alignment, but Prior NRC Approval.............

- 49 3.2.1.4 Attribute Alignment - Not in Alignment, but No Adverse Consequences.

- 49 3.2.1.5 Attribute Alignment - Not in Alignment....................................................

- 49 3.2.1.6 NFPA 805 NSCA Methods Conclusion..................... :..............................

- 49 3.2.2 Maintaining Fuel in a Safe and Stable Condition,..................................................

- 49

. 3.2.3 Applicability of Feed and Bleed..............................................................................

- 50 3.2.4 Assessment of Multiple Spurious Operations.........................................................

- 50 3.2.5 Establishing Recovery Actions................................................................................

- 51 3.2.6 Conclusion for Section 3.2.....................................................................................

- 53 3.3 Fire Modeling................................................. *............................................ :.............

- 54 3.4 Fire Risk Assessments..................................................... :....................................

- 54 3.4.1 Maintaining Defense-in-Depth and Safety Margins................................................

- 55 3.4.1.1 Defense-in-Depth (DID)..........................................................................

- 55 3.4.1.2 Safety Margins.................... :................................................................... - 56 3.4.1.3 Defense-in-Depth and Safety Margin Conclusion....................................

- 57

- iii 3.4.2 Quality of the Fire Probabilistic Risk Assessment.....................................

- 58 3.4.2.1 Internal Events PRA Model......................................................................

- 58 3.4.2.2 Fire PRA Model........................................................................................

- 59 3.4.2.3 Fire Modeling in Support of the Development of the Fire Risk Evaluations (FREs)...................................................................................

- 65 3.4.2.3.1 Overview of Fire Models Used to Support the FREs..........................

- 66 3.4.2.3.2 RAls Pertaining to Fire Modeling in Support of the DAEC Fire PRA..

- 68 3.4.2.3.3Conclusion for Section 3.4.2.3...........................................................

- 71 3.4.2.4 Conclusions Regarding Fire PRA Quality................................................

- 71 3.4.3 Fire Risk Evaluations................................................................................... ;.........

- 72 3.4.4 Additional Risk Presented by Recovery Actions....................................................

- 75 3.4.5 Risk-Informed or Performance-Based Alternatives to Compliance with NFPA 805

- 75 3.4.6 Cumulative Risk and Combined Changes.............................................................

- 76 3.4.7 Uncertainty and Sensitivity Analyses.....................................................................

- 77 3.4.8 Conclusion for Section 3.4.................... ;................................................................

- 78 3.5 Nuclear Safety Capability Assessment Results.....................................................-

- 79 3.5.1 Nuclear Safety Capability Assessment Results by Fire Area.................................

- 80 3.5.1.1 Fire Detection and Suppression Systems Required to meet the NSPC...

- 82 3.5.1.2 Evaluation of Fire Suppression Effects on NSPC.....................................

- 83 3.5.1.3 Licensing Actions.....................................................................................

- 83 3.5.1.4 Existing Engineering Equivalency Evaluations (EEEEs)...........................

- 86 3.5.1.5 Variances from Deterministic Requirements....................,....... :...............

- 86 3.5.1.6 Recovery Actions.....................................................................................

- 87

.3.5.1.7 RAs Credited for Defense in Depth..........................................................

- 87 3.5.1.8 Plant Fire Barriers and Separations.........................................................

- 87 3.5.1.9 Electrical Raceway Fire Barrier Systems'..................................................

- 88 3:5.1.10 Conclusion for Section 3.5.1....................................................................

- 88 3.5.2 Clarification of Prior NRC Approvals......................................................................

- 89 3.5.3 Fire Protection during Non-Power Operational Modes,(NPO) Modes....................

- 89 3.5.3.1 !\\IPO Strategy and Plant Operating States (POSs)...................................

- 89 3.5.3.2 !\\IPO Analysis Process.............................................................................

- 90 3.5.3.3 NPO Key Safety Functions and SSCs Used to Achieve Performance....

- 91

- iv 3.*5.3.4 NPO Pinch Point Resolutions and Program Implementation....................

- 92 3.5.4 Conclusion for Section 3.5.....................................................................................

- 93 3.6 Radioactive Release Performance Criteria............................................................

- 94 3.6.1 Gaseous Effluent Controls.....................................................................................

- 96 3.6.2 Liquid Effluent Controls.........................................................................................

- 97 3.6.3 Pre-Fire Plans.......................................................................................................

- 98 3.6.4 Fire Brigade Training Programs.............................................................................

- 98 3.6.5 Actions to be Completed........................................................................................

- 99 3.6.6 Conclusion for Section 3.6.....................................................................................

- 99 3.7 NFPA 805 Monitoring Program............................................................................

- 100 3.7.1 Conclusion for Section 3.7...................................................................................

-101 3.8 Program Documentation, Configuration Control, and Quality Assurance.............

- 101 3.8.1 Documentation....................................................................................................

- 102 3.8.2 Configuration Control............................................................................................. - 102 3.8.3 Fire Modeling Quality............................................................................................. - 103 3.8.3.1 Review.................................................................................................... - 103 3.8.3.2 Verification and Validation (V&V)............................................................ - 103 3.8.3.2.1 GeneraL............................................................................................. - 104 3.8.3.2.2 Discussion of Selected RAI Responses............................................. - 105 3.8.3.2.3 Post-Transition.................................................................................. - 106 3.8.3.2.4 Conclusion for Section 3.8.3.2........................................................... - 106 3.8.3.3 Limitations of Use.....................,............................................................. - 106 3.8.3.3.1 GeneraL............................................................................................. - 106 3.8.3.3.2 Discussion of RAls............................................................................. - 107 3.8.3.3.3 Post-Transition.................................................................................. - 108 3.8.3.3.4 Conclusion for Section 3.8.3.3........................................................... - 108 3.8.3.4 Qualification of Users.............................................................................. - 108 3.8.3.5 Uncertainty Analysis................................................................................ - 110 3.8.3.5.1 GeneraL.............................................................................................. - 110 3.8.3.5.2 Discussion of Fire Modeling RAls...................................................... - 111 3.8.3.5.3 Post-Transition................................................................................. : - 111 3.8.3.5.4 Conclusion for Section 3.8.3.5........................................................... - 112

- v 3.8.3.6 Conclusion for Section 3.8.3................................................................... - 112 3.8.4 Fire Protection Quality Assurance Program....................... :................................... - 112 3.8.5 Conclusion for Section 3.8..................................................................................... - 112 4.0 FIRE PROTECTION LICENSE CONDITION......................................................... -113 5.0 SUMMARy............................................................................................................ -115

6.0 STATE CONSULTATION

..................................................................................... -116

7.0 ENVIRONMENTAL CONSIDERATION

............... ~................................................. -116

8.0 CONCLUSION

........................................................................... ~.......................... -116

9.0 REFERENCES

......... ~.............................................................................................- 118 10.0 ABBREVIATIONS AND ACRONyMS................................................................... - 127 ATTACHMENTS Attachment A: Table 3.8-1, V&V Basis for Fire Modeling Correlations Used at DAEC.................................................................................................. - A1 Attachment B: Table 3.8-2, V&V Basis for Fire Model Calculations of Other Models Used at DAEC...........................................................................

- B1

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BYTHE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 286 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-49 TRANSITION TO A PERFORMANCE~BASED FIRE PROTECTION PROGRAM IN ACCORDANCE WITH 10 CFR 50.48(c)

NEXTERA ENERGY DUANE ARNOLD. LLC DUANE ARNOLD ENERGY CENTER DOCKET NO. 50-331

1.0 INTRODUCTION

1.1 Background

The NRC started developing fire protection requirements in the 1970s, and in 1976, the NRC published comprehensive fire protection guidelines. Subsequently, the NRC performed fire protection reviews for the operating reactors, and documented the results in safety evaluation reports (SERs) or supplements to SERs. In 1980, to resolve issues identified in those reports, the NRC amended its regulations for fire protection in operating nuclear power plants and published its Final Rule, Fire Protection Program for Operating Nuclear Power Plants, 45 Fed.

Reg. 76,602 (Nov. 19, 1980) (adding 10 CFR 50.48, "Fire protection" (Reference 1) and Appendix R (Reference 4) to 10 CFR Part 50 "Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979). Section 50.48(a)(1) requires each operating nuclear power plant to have a fire protection plan that satisfies General Design Criterion (GDC) 3 of Appendix A to 10 CFR 50 (Reference 2 and Reference 3) and states that the fire protection plan must describe the overall fire protection program; identify the positions responsible for the program and the authority delegated to those positions; outline the plans for fire protection, fire detection and suppression capability, and limitation of fire damage. Section 50.48(a)(2) states that the fire protection plan must describe the specific features necessary to implement the program described in paragraph (a)(1) including administrative controls and personnel requirements; automatic and manual fire detection and suppression systems; and the means to limit fire damage to structures, systems, and components (SSCs) to ensure the capability to safely shut down the plant. Section 50.48(a)(3) requires that the licensee retain the fire protection plan and each 'change to the plan as a record until the Commission terminates the license.

In the 1990s, the NRC worked with the National Fire Protection Association (NFPA) and industry to develop a risk-informed (RI), performance-based (PB) consensus standard for fire protection. In 2001, the NFPA Standards Council issued NFPA 805, whi9h describes a

- 2 methodology for establishing fundamental fire protection program design requirements and elements, determining required fire protection systems and features, applying performance based requirements, and administering fire protection for existing light water reactors during operation, decommissioning, and permanent shutdown. It provides for the establishment of a minimum set of fire protection requirements but allows performance-based or deterministic approaches to be used to meet performance criteria.

Regulatory Guide 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light Water Nuclear Power Plants," Revision 1, (RG 1.205) (Reference 8), states:

On March 26, 1998, the NRC staff sent to the Commission SECY-98-058, "Development of a Risk-Informed [RI], Performance-Based Regulation for Fire Protection at Nuclear Power Plants" [Reference 5], in which it proposed to work with the NFPA and the industry to develop a risk informed, performance based

[RI/PB] consensus standard for nuclear power plant fire protection. This consensus standard could be endorsed in a future rulemaking as an alternative set of fire protection requirements to the existing regulations in 10 CFR 50.48. In SECY-00-0009, "Rulemaking Plan, Reactor Fire Protection Risk-Informed, Performance-Based Rulemaking," dated January 13, 2000 [Reference 6], the NRC staff requested and received Commission approval to proceed with rulemaking to permit operating reactor licensees to adopt an NFPA standard as an alternative to existing fire protection requirements. On February 9, 2001, the NFPA Standards Council approved the 2001 edition of NFPA 805,

["Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," [Reference 7)) as an American National Standard

[ANS] for performance-based [PB] fire protection for light-water nuclear power plants.

An adoptee of NFPA 805 must meet the performance goals, objectives, and criteria that are itemized in Chapter 1 of NFPA 805 through the implementation of performance-based or deterministic approaches. The goals include ensuring that reactivity control, inventory and pressure control, decay heat removal, vital auxiliaries, and process monitoring are achieved and maintained. The adoptee then must establishes plant fire protection requirements using the methodology in Chapter 2 of NFPA 805 such that the minimuhl fire protection program elements and design criteria contained in Chapter 3 of NFPA 805 are satisfied. Next, an adoptee identifies fire areas and fire hazards though a plant-wide analysis, and then applies either a performance-based or a deterministic approach to meet the performance criteria. As part of a performance-based approach, an adoptee will use engineering evaluations, probabilistic safety assessments, and fire modeling calculations to show that the criteria are met. Chapter 4 of NFPA establishes the methodology to determine the fire protection systems and features required to achieve the performance criteria. It also specifies that at least one success path to achieve the nuclear safety performance criteria shall be maintained free of fire damage by a single fire.

RG 1.205 also states that:

Effective July 16, 2004, the Commission amended its fire protection requirements in 10 CFR 50.48 to add 10 CFR 50.48( c), which incorporates by reference the

- 3 2001 Edition of NFPA 805, with certain exceptions, and allows licensees to apply for a license amendment to comply with the 2001 edition of NFPA 805 (69 FR 33536). NFPA has issued subsequent editions of NFPA 805, but the regulation does not endorse them.

Throughout this safety evaluation (SE), where the NRC staff states that the licensee's FPP element is in compliance with (or meeting the requirements of) NFPA 805, the NRC staff is referring to NFPA 805 with the exceptions, modifications, and supplementation described in 10 CFR 50.48( c )(2).

RG 1.205 also states that:

, In parallel with the Commission's efforts to issue a rule incorporating the risk informed, performance-based fire protection provisions of NFPA 805, [the Nuclear Energy Institute] (NEI) published implementing guidance for the specific provisionsof NFPA 805 and 10 CFR 50.48(c) in NEI 04-02, ["Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c)".]

Regulatory Guide (RG) 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1, (RG 1.205) provides the NRC staff's position on NEI 04-02, Revision 2 (NEI 04-02) [Reference 9], and offers additional information and guidance to supplement th~ NEI document and assist licensees in meeting the NRC's regulations in 10 CFR 50.48(c) related to adopting a RIIPB FPP.

Accordingly, NextEra Energy Duane Arnold, LLC (NextEra or the licensee), requested a license amendment to allow the licensee to maintain the Duane Arnold Energy Center (DAEC) FPP in accordance with 10 CFR 50.48( c).

1.2 Requested Licensing Action By application sent to the U.S. NRC dated August 5,2011 (Reference 10), as supplemented by letters dated October 14, 2011, April 23, 2012, May 23, 2012, July 9, 2012, October 15, 2012, January 11, 2013, February 12, 2013, March 6, 2013, May 1, 2013, May 29, 2013, two supplements dated July 2, 2013, August 5, 2013, and August 28, 2013 (References 11, 12, 13, 14, 15, 16, 17, 18, 19, 20, 21, 22, 93, and 95), NextEra Energy Duane Arnold, LLC submitted an application for a license amendment to transition the DAEC FPP from 10 CFR 50.48(b), to 10 CFR 50.48(c), NFPA 805, "Performance-Based Standard for Fire Protection For Light Water Reactor Electric Generating Plants," 2001 Edition. The supplements provided additional information that clarified the application, but did not expand the overall scope of the application as originally noticed, and did not change the NRC staffs original proposed opportunity for a hearing on the initial application as published in the Federal Register (FR) on October 2, 2012 (77 FR 60151).

The licensee, requested an amendment to the DAEC renewed operating license and technical specifications (TSs) in order to establish and maintain a RI/PB FPP in accordance with the requirements of 10 CFR 50.48(c).

- 4 Specifically, the licensee requested to transition from the existing deterministic fire protection licensing basis established in accordance with Final Safety Analysis Report and as approved in the Safety Evaluation Report (SER) dated June 1, 1978 (Reference 23) and supplement dated February 10,1981 (Reference 24), to a PB FPP in accordance with 10 CFR 50.48(c), that uses risk information, in part, to demonstrate compliance with the fire protection and nuclear safety goals, objectives, and performance criteria of NFPA 805. As such, the proposed FPP at DAEC is referred to as RI/PB throughout this Safety Evaluation (SE).

In its LAR, the licensee has provided a description of the revised fire protection program for which it is requesting NRC approval to implement, a description of the FPP that it will implement under 10 CFR 50.48(a)and (c), and the results of the evaluations and analyses required by NFPA 805.

This SE documents the NRC staff's evaluation of the licensee's LAR and the NRC staff's conclusion that:

(1) The licensee has identified any orders and license conditions that must be revised or superseded, and has provided the necessary revisions to the plant's TSs and bases, as required by 10 CFR 50.48(c)(3)(i).

(2) The licensee has completed its implementation of the methodology in Chapter 2, "Methodology," of NFPA 805 (including all required evaluations and analyses), and the NRC staff has approved the licensee's modified fire protection plan, which reflects the decision to comply with NFPA 805, as required by 10 CFR 50.48(a).

(3) The licensee will modify its FPP, as described in the LAR, in accordance with the implementation schedule set forth in this SE and the accompanying license condition, as required by 10 CFR 50.48(c)(3)(ii).

The licensee proposed a new fire protection license condition reflecting the new RI/PB FPP licensing basis. Section 2.4.2 and Section 4.0 of this SE discuss the license condition in detail and Section 2.4.3 discusses TS changes.

2.0 REGULATORY EVALUATION

Section 50.48, "Fire Protection," of 10 CFR(Reference 1) provides the NRC requirements for nuclear power plant fire protection.

The NRC regulations include specific requirements for requesting approval for a RI/PB FPP based on the provisions of NFPA 805 (Reference 7). Paragraph 50.48(c)(3)(i) of 10 CFR states, in part, that:

A licensee may maintain a FPP that complies with NFPA 805 as an alternative to complying with [10 CFR 50.48(b)] for plants licensed to operate before January 1, 1979, or the fire protection license conditionsJorplants licensed to operate after January 1, 1979. The licensee shall submit a request to comply with NFPA 805 in the form of an application for license amendment under [10 CFR] 50.90.

"The application must identify any orders and license conditions that must be

- 5 revised or superseded, and contain any necessary revisions to the plant's technical specifications and the bases thereof.

In addition, 10 CFR 50.48(c)(3)(ii) states that:

The licensee shall complete its implementation of the methodology in Chapter 2 of NFPA 805. (including all required evaluations and analyses) and, upon completion, modify the fire protection plan required by paragraph (a) of this section to reflect the licensee's decision to comply with NFPA 805, before changing its FPP or nuclear power plant as permitted by NFPA 805.

The intent of 10 CFR 50.48( c)(3)(ii) is given in the statement of considerations for the Final Rule, Voluntary Fire Protection Requirements for Light Water Reactors; Adoption of NFPA 805 as a Risk-Informed, Performance-Based Alternative, 69 Fed. Reg. 33,536, 33,548 (June 16, 2004), which states:

This paragraph requires licensees to complete all of the Chapter 2 methodology (including evaluations and analyses) and to modify their fire protection plan before making changes to the FPP or to the plant configuration. This process ensures that the transition to an NFPA 805 configuration is conducted in a complete, controlled, integrated, and organized manner. This requirement also precludes licensees from implementing NFPA 805 on a partial or selective basis (e.g., in some fire areas and not others, or truncating the methodology within a given fire area).

As stated in 10 CFR 50.48( c )(3)(i), the Director of the Office of Nuclear Reactor Regulation (NRR), or a designee of the Director, may approve the application if the Director or designee determines that the licensee has identified orders, license conditions, and the technical specifications that must be revised or superseded, and that any necessary revisions are adequate.

The regulations also allow for flexibility that was not included in the NFPA 805 standard.

Licensees who choose to adopt 10 CFR 50.48(c), but wish to use the PB methods permitted elsewhere in the standard to meet the fire protection requirements of NFPA 805 Chapter 3, "Fundamental Fire Protection Program and Design Elements," may do so by submitting aLAR in accordance with 10 CFR 50.48( c)(2)(vii).

The Director of the Office of Nuclear Reactor Regulation, or a designee of the Director, may approve the application if the Director or designee determines that the performance-based approach; (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

- 6 (C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).

Alternatively, licensees may choose to use RI or PB alternatives to comply with NFPA 805 by submitting a LAR in accordance with 10 CFR 50.48(c)(4).

The Director of the Office of Nuclear Reactor Regulation, or designee of the Director, may approve the application if the Director or designee determines that the proposed alternatives:

(i)

Satisfy the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological*

release; (ii)

Maintain safety margins; and (iii) Maintain fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).

In addition to the conditions outlined by the rule that require licensees to submit a LAR for NRC review and approval in order to adopt a RI/PB FPP, a licensee may also submit additional elements of its FPP for which it wishes to receive specific NRC review and approval, as set forth in Regulatory Position C.2.2.1 of RG 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1 (Reference 8). Inclusion of these elements in the NFPA 805 LAR is meant to alleviate uncertainty in portions of the current FPP licensing bases as a result of the lack of specific NRC approval of these elements. Regulatory guides are not substitutes for regulations, and compliance with them is not required. Methods and solutions that differ from those set forth in regulatory guides will be deemed acceptable if they provide a basis for the findings required for the issuance or continuance o1'a permit or license by the Commission. Accordingly, any submittal addressing these additional FPP elements needs to include sufficient detail to allow the NRC staffto assess whether the licensee's treatment of these elements meets the 10 CFR 50.48(c) requirements The purpose of the FPP established by NFPA 805 is to provide assurance, through a defense in-depth (DID) philosophy, that the NRC's fire protection objectives are satisfied. NFPA 805 Section 1.2, "Defense-in-Depth," states the following:

Protecting the safety of the public, the environment, and plant personnel from a plant fire and its potential effect on safe reactor operations is paramount to this standard. The fire protection standard shall be based on the concept of defense-in-depth. Defense-in-depth shall be achieved when an adequate balance of each of the following elements is provided:

(1)

Preventing fires from starting; (2)

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

- 7 (3)

Providing an adequate level of fire protection for structures, systems and components important to safety, so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed In addition, in accordance with GDC 3, "Fire protection," of Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, fire protection systems must be designed such that their failure or inadvertent operation does not significantly impair the ability of the structures, systems and components (SSCs) important to safety to perform their intended safety functions.

2.1 Applicable Regulations The following regulations address fire protection:

GDC 3, "Fire protection," to 10 CFR Part 50, Appendix A:

Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions.

Noncombustible and heat resistant materials shall be used wherever practical throughout the unit, particularly in locations such as the containment and control room. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety.

Firefighting systems shall be designed.to assure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components.

GDC 5, "Sharing of structures, systems, and components" to 10 CFR Part 50, Appendix A, relates to shared fire protection systems, and potential fire impacts on shared SSCs important to safety.

10 CFR 50.48(a)(1), requires that each holder of an operating license have a fire protection plan that satisfies General Design Criterion 3 of appendix A to 10 CFR Part 50.

10 CFR 50.48(c), incorporates t\\lFPA 805 (2001 Edition) by reference, with certain exceptions, modifications and supplementation. This regulation establishes the requirements for using a RI/PB FPP in conformance with NFPA 805 as an alternative to the requirements associated with 10 CFR 50.48(b) and Appendix R, "Fire Protection Program for Nuclear Power Facilities Operating Prior to JClnuary 1, 1979," to 10 CFR Part 50, or the specific plant fire protection license condition.

10 CFR Part 20, "Standards for Protection Against Radiation," establishes the radiation protection limits used as NFPA 805 radioactive release performance criteria, as specified in NFPA 805, Section 1.5.2, "Radioactive Release Performance Criteria."

- 8 2.2 Applicable Staff Guidance The NRC staff review also relied on the following additional codes, RGs, and standards:

RG 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1, issued December 2009, (Reference 8), provides guidance for use in complying with the requirements that the NRC has promulgated for RI/FP FPPs that comply with 10 CFR 50.48 and the referenced 2001 Edition of the NFPA standard. It endorses portions of NEI 04-02, Revision 2, where it has been found to provide methods acceptable to the NRC for implementing NFPA 805 and complying with 10 CFR 50.48(c). The regulatory positions in Section C of RG 1.205 include clarification of the guidance provided in NEI 04-02, as well as NRC exceptions to the guidance. RG 1.205 sets forth regulatory positions, emphasizes certain issues, clarifies the requirements of 10 CFR 50.48( c) and NFPA 805, clarifies the guidance in NEI 04-02, and modifies the NEI 04-02 guidance where required. Should a conflict occur between NEI 04-02 and this RG, the regulatory positions in RG 1.205 govern.

• The 2001 edition of NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reador Electric Generating Plants," which specifies the minimum fire protection requirements for existing light water nuclear power plants during all phases of plant operations, including shutdown, degraded conditions, and decommissioning.

NFPA 805 was developed to prqvide a comprehensive risk-informed, performance based standard for fire protecti6ri. The NFPA' 805 Technical Committee on Nuclear Facilities is composed of nuclear plant licensees, the NRC, insurers, equipment manufacturers, and subject matter experts. The standard was developed in accordance with NFPA processes, and consisted of a number of technical meetings and reviews of draft documents by committee and industry representatives. The scope of NFPA 805 includes goals related to nuclear safety, radioactive release, life safety, and plant damage/business interruption. The standard addresses fire protection requirements for nuclear plants during all plant operating modes and conditions, including shutdown and decommissioning, which had not been explicitly addressed by previous requirements and guidelines: NFPA 805 became effective on February 9,2001.

NEI 04-02 "Guidance for Implementing a Risk-Informed, Performance~Based Fire Protection Program Under 10 CFR 50.48(c),"which provides guidance for implementing the requirements of 10 CFR 50.49(c), and represents methods for implementing in whole or in part a risk-informed, performance-based fire protection program. This implementing guidance for NFPA 805 has two primary purposes: (1) provide direction and clarification

, for adopting NFPA 805 as an acceptable approach to fire protection, consistent with 10 CFR 50.48 (c); and (2) provide additional supplemental technical guidance and methods for using NFPA 805 and its appendices to demonstrate compliance with fire protection requirements. Although there is a significant amount of detail in NFPA 805 and its appendices, clarification and additional guidance for select issues help ensure consistency and effective utilization of the standard. The NEI 04-02 guidance focuses attention on the risk-informed, performance-based fire protection goals, objectives, and performance criteria contained in NFPA 805 and the risk-informed, performance-based tools considered acceptable for demonstrating compliance. Revision 2 of NEI 04-02

- 9 incorporates guidance from RG 1.205 and approved Frequently Asked Questions (FAQs)

RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to(the Licensing Basis," Revision 2, issued May 2011, (Reference 25), which provides the NRC staff's recommendations for using risk information in support of licensee-initiated licensing basis (LB) changes to a nuclear power plant that require such review and approval. The guidance provided does not preclude other approaches for requesting LB changes. Rather, RG 1.174 is intended to improve consistency in regulatory decisions in areas in which the results of risk analyses are used to help justify regulatory action. As such, the RG provides general guidance concerning one approach that the NRC has determined to be acceptable for analyzing issues associated with proposed changes to a plant's LB and for assessing the impact of such proposed changes on the risk associated with plant design and operation.

RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, issued March 2009, (Reference 26), which provides guidance to licensees for use in determining the technical adequacy of the base probabilistic risk assessment (PRA) used in a risk informed regulatory activity, and endorses standards and industry peer review guidance.

The RG provides guidance in four areas:

(1) a definition of a technically acceptable PRA (2) the NRC's position on PRA consensus standards and industry PRA peer review program documents (3) demonstration that the baseline PRA (in total or specific pieces) used in regulatory applications is of sufficient technical adequacy (4) documentation to support a regulatory submittal It does not provide guidance on how the base PRA is revised for a specific application or how the PRA results are used in application-specific decision-making processes.

RG 1.189, "Fire Protection for Operating Nuclear Power Plants," Revision 2, issued October 2009, (Reference 27), which provides guidance to licensees on the proper content and quality of engineering equivalency evaluations used to support the FPP.

The NRC staff developed the RG to provide a comprehensive fire protection guidance document and to identify the scope and depth of fire protection that the staff would consider acceptable for nuclear power plants.

NUREG 0800, Section 9.5.1.2, "Risk-Informed, Performance-Based Fire Protection Program," Revision 0, issued December 2009, (Reference 28), which provides the NRC staff with guidance for evaluating LARs that seek to implement a RI/PB FPP in accordance with 10 CFR 50.48(c).

- 10 NUREG 0800, Section 1.9.1, "Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 3, issued September 2012, (Reference 29), which provides the NRC staff with guidance for evaluating the technical adequacy of a licensee's PRA results when used to request RI changes to the licensing basis.

NUREG 0800, Section 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance," Revision 0, issued June 2007, (Reference 30), which provides the NRC staff with guidance for evaluating the risk information used by a licensee to support permanent RI changes to the licensing basis.

NUREG/CR-6850, "EPRIlf\\lRC-RES Fire PRA Methodology for Nuclear Power Facilities," Volumes 1 and 2, (Reference 62), which presents a compendium of methods; data and tools to perform a fire probabilistic risk assessment and develop associated insights. In order to address the need for improved methods, the NRC Office of Nuclear Regulatory Research,(RES) and Electric Power Research Institute (EPRI) embarked upon a program to develop state-of-art Fire PRA methodology. Both RES and EPRI have provided specialists in fire risk analysis, fire modeling,electrical engineering, human reliability analysis, and systems engineering for methods development. A formal technical issue resolution process was developed to direct the deliberative process between RES and EPRI. The process ensures that divergent technical views are fully considered, yet encourages consensus at many points during the deliberation.

Significantly, the process provides that each party maintain its own point of view if consensus is not reached. Consensus was reached on all technical issues documented in NUREG/CR-6850.. The methodology documented in this report reflects the current state-of-the-art in Fire PRA. These methods are expected to form a basis for risk informed analyses related to the plant fire protection program. Volume 1, the Executive Summary, provides general background and overview information including both

. programmatic and technical, and project insights and conclusions. Volume 2 provides the detailed discussion of the recommended approach, methods, data and tools for conduct of a Fire PRA.

The NRC staff notes that, based on new experimental information, the reduction in hot short probabilities for circuits provided with control power transformers (CPT) identified in NUREG/CR 6850 cannot be repeated in experiments and therefore may be too high and should be reduced.

NUREG-1805, "Fire Dynamics Tools (FDTs): Quantitative Fire Hazard Analysis Methods for the U.S. Nuclear Regulatory Commission Fire Protection Inspection Program," (Reference 67) which provides quantitative methods, known as "Fire Dynamics Tools" (FDTs), to assist regional fire protection inspectors in performing fire hazard analysis. The FDTs are intended to assist fire protection inspectors in performing risk-informed evaluations of credible fires that may cause critical damage to essential safe-shutdown equipment, as required by the new reactor oversight process defined in the NRC's inspection manual.

- 11 NliREG-1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications;" Volumes 1 through 7, (Reference 68) which provide technical documentation regarding the predictive capabilities of a specific set of fire models for the analysis of fire hazards in nuclear power plant (NPP) scenarios. This report is the result of a collaborative program with the Electric Power Research Institute (EPRI) and the National Institute of Standards and Technology (NIST)The selected models are:

(1) FDTs developed by NRC (Volume 3)

(2) FIVE-Rev1 developed by EPRI (Volume 4)

(3) The zone model CFAST developed by NIST (Volume 5)

(4) The zone model MAGIC developed by Electricite de France (EdF)

(Volume 6)

(5) The computational fluid dynamics model FDS developed by NIST (Volume 7).

In addition to the fire model volumes, Volume 1 is the comprehensive main report and Volume 2 is a description of the experiments and associated experimental uncertainty used in developing this report.

NUREG/CR-7010, "Cable Heat Release, Ignition, and Spread In Tray Installations during Fire {CHRISTI FIRE), Volume 1: Horizontal Trays," (Reference 83), which describes Phase 1 of the CHRISTIFIRE testing program conducted by NIST. The overall goal of this multiyear program is to quantify the burning characteristics of grouped electrical cables installed in cable trays. This first phase of the program focuses on horizontal tray configurations. CHRISTI FIRE addresses the burning behavior of a cable in a fire beyond the point of electrical failure. The data obtained from this project can be used for.

the development of fire models to calculate the heat release rate (HRR) and flame spread of a cable fire.

• NUREG-1855, Volume 1, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," (Reference 89), which provides guidance on how to treat uncertainties associated with PRA in risk-informed decision-making. The objectives of this guidance include fostering an understanding of the uncertainties associated with PRA and their impact on the results of PRA and providing a pragmatic approach to addressing these uncertainties in the context of the decision-making. To meet the objective of the NUREG, it is necessary to understand the role that PRA results play in the context of the decision process. To define this context, NUREG-1855 provides an overview of the risk-informed decision-making process itself.

Generic Letter (GL) 2006-03. "Potentially Nonconforming Hemyc and MT Fire Barrier Configurations," (Reference 49), which requested that licensees evaluate their facilities to confirm compliance with the existing applicable regulatory requirements in light of the information provided in this GL and, if appropriate, take additional actions.

NEI 00-01, "Guidance for Post Fire Safe Shutdown Circuit Analysis," Revision 2, (Reference 55), which provides a deterministic methodology for performing post-fire safe shutdown analysis. In addition, NEI 00-01 includes information on risk-informed

- 12 methods (when allowed within a Plant's License Basis) that may be used in conjunction with the deterministic methods for resolving circuit failure issues related to Multiple Spurious Operations (MSOs). The risk-informed method is intended for application by licensees to determine the risk significance of identified circuit failure issues related to MSOs.

iii ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," (Reference 59), which provides guidance PRAs used to support risk-informed decisions for commercial light water reactor nuclear power plants and prescribes a method for applying these requirements for specific applications. The Standard gives guidance for a Level 1 PRA of internal and external hazards for all plant operating modes. In addition, the Standard provides guidance for a limited Level 2 PRA sufficient to evaluate large early release frequency (LERF). The only hazards explicitly excluded from the scope are accidents resulting from purposeful human-induced security threats (e.g., sabotage). The Standard applies to PRAs used to support applications of risk-informed decision-making related to design, licensing, procurement, construction, operation, and maintenance.

2.3 NFPA 805 Frequently Asked Questions In the LAR, the licensee proposed to use a number of documents commonly known as NFPA 805 FAQs. The following table provides the set of FAQs the licensee used that the NRC staff referenced in the preparation of this SE, as well as the SE section(s) to which each FAQ was referenced.

Table 2.3-1: NFPA 805 Frequently Asked Questions FAQ#

FAQ Title and Summary Ref.

SE Section 07-0030 "Establishing Recovery Actions"

• This FAQ provides an acceptable process for determining th-e recovery actions for NFPA 805 Chapter 4 compliance.

The process includes:

Differentiation between recovery actions and activities in the main control room or at primary control station( s).

Determination of which recovery actions are required by the NFPA 805 fire protection program.

Evaluate the additional risk presented by the use of recovery actions.

Evaluate the feasibility of the identified recovery actions.

Evaluate the reliability of the identified recovery actions.

31 3.2.5 07-0038 "Lessons Learned on Multiple Spurious Operations (MSOs)"

32 3.2.1, This FAQ reflects an acceptable process for the treatment of MSOs during transition to NFPA 805:

Step 1 - Identify potential MSO combinations of concern.

3.2.4

- 13 Step 2 - Expert panel assesses plant specific vulnerabilities and reviews MSOs of concern.

Step 3 - Update the fire PRA and Nuclear Safety Capability Assessment to include MSOs of concern.

Step 4 - Evaluate for NFPA 805 compliance.

Step 5 - Document the results.

07-0039

""Incorporation of Pilot Plant Lessons Learned - Table B-2" This FAQ provides additional detail for the comparison of the licensee's safe shutdown strategy to the endorsed industry guidance, NEI 00-01 "Guidance for Post-Fire Safe Shutdown Circuit Analysis," Revision 1 (Reference 55). In short, the process has the licensees:

Assemble industry and plant-specific documentation; Determine which sections of the guidance are applicable; Compare the existing safe shutdown methodology to the applicable guidance; and Document any discrepancies.

33 3.2.1 07-0040 "Non-Power Operations (NPO) Clarifications" This FAQ clarifies an acceptable NFPA 805 NPO program.

The process includes:

Selecting NPO equipment and cabling.

Evaluation of NPO Higher Risk Evolutions (HRE).

Analyzing NPO key safety functions (KSF).

Identifying plant areas to protect or "pinch points" during NPO HREs and actions to be taken if KSFs are lost.

34 3.5.3, 3.5.3.1 thru 3.5.3.4 08-0048 "Revised Fire Ignition Frequencies" This FAQ provides an acceptable method for using updated fire ignition frequencies in the licensee's fire PRA. The method involves the use of sensitivity studies when the updated fire ignition frequencies are used.

35 3.4.6, 3.4.7 08-0054 "Compliance with Chapter 4 of NFPA 805" This FAQ provides an acceptable process to demonstrate Chapter 4 compliance for transition:

Step 1 - Assemble documentation Step 2 - Document Fulfillment of Nuclear Safety Performance Criteria Step 3 - Variance From Deterministic Requirements (VFDR) Identification, Characterization, and Resolution Considerations Step 4 - Performance-Based Evaluations Step 5 - Final VFDR Evaluation Step 6 - Document Required Fire Protection Systems 36 3.5.1.4

- 14 and Features 09-0056 "Radioactive Release Transition" This*FAQprovides an acceptable level of detail and content for the radioactive release section of the LAR. It includes:

Justification of the compartmentation, if the radioactive release review is not performed on a fire area basis.

Pre-fire plan and fire brigade training review results.

Results from the review of engineering controls for gaseous and liquid effluents.

37 3.6 10-0059 "Monitoring Program" This FAQ provides clarification regarding the implementation of an I\\IFPA 805 monitoring program for transition. It includes:

Monitoring program analysis units;

)

Screening of low safety significant structures, systems, and components; Action level thresholds; and The use of existing monitoring programs.

38 3.7 12-0062 "Updated Final Safety Analysis Report (UFSAR) Content" This FAQ provides.the. necessary level of detail for the transition of the fire protection sections within the UFSAR.

39 2.4.4 2.4 Orders, License Conditions, and Technical Specifications Paragraph 50.48(c)(3)(i) of 10 CFR states that the LAR "". must identify any orders and license conditions that must be revised or superseded, and contain any necessary revisions to the plant's technical specifications and the bases thereof."

2.4.1 Orders The NRC staff reviewed Section 5.2.3, "Orders and Exemptions" and Attachment 0, !'Orders and Exemptions" of DAEC's LAR, with regard to !\\IRC-issued orders pertinent to DAEC that are being revised or superseded by the NFPA 805 transition process. The LAR stated that the licensee conducted a review of NextEra docketed correspondence to determine if there were any orders or exemptions that needed to be superseded or revised. The LAR also stated that the licensee conducted a review to ensure that compliance with the physical protection requirements, security orders, and adherence to those commitments applicable to DAEC are maintained. The licensee discussed the affected orders and exemptions in Attachment O. The licensee requested that 13 exemptions be rescinded, and determined that no orders need to be superseded or revised to implement a FPP at DAEC that complies with 10 CFR 50.48(c).

This review, conducted by NextEra, included an assessment of docketed correspondence files and electronic searches, including the NRC's Agencywide Document Access and Management

- 15 System (ADAMS) document management system. The review was performed to ensure that compliance with the physical protection requirements, security orders, and adherence to commitments applicable to DAEC are maintained. The NRC staff accepts the licensee's determination that 13 exemptions are rescinded and another 3 exemptions are rescinded but transitioned to NFPA 805 as listed in Attachment K of the LAR and that no Orders need to be superseded or revised to implement NFPA 805 at DAEC. See Section 2.5 of this SE for the NRC staff's detailed evaluation of the exemptions being rescinded.

In addition, the licensee performed a specific review of the license amendment that incorporated the mitigation strategies required by 10 CFR 50.54(hh)(2) to ensure that any changes being made in order to comply with 10 CFR 50.48(c) do not invalidate existing commitments applicable to DAEC. The licensee's review of this regulation and the related license amendment demonstrated that changes to the FPP during transition to NFPA 805 will not affect the mitigation measures required by 10 CFR 50.54(hh )(2). The !\\IRC staff accepts the licensee's determination in regard to 10 CFR 50.54(hh)(2).

2.4.2 License Conditions The NRC staff reviewed LAR Section 5.2.1, "License Condition Changes," and Attachment M, "License Condition Changes," regarding changes the licensee seeks to make to the DAEC*fire protection license condition in order to adopt NFPA ~05, as required by 10 CFR 50.48(c)(3).

The NRC staff reviewed th,e revised license conditions, which replaces the current DAEC fire protection license condition 2.C(3), for consistency with the format and content guidance in Regulatory Position C.3.1 of RG 1.205, Revision 1, and with the proposed plant modifications identified in the LAR.

The license conditions provide a structure and detailed criteria to allow self-approval for RI/PB as well as other types of changes to the FPP. The structure and detailed criteria re_sult in a process that meets the requirements in Sections 2.4, Engineering Analyses, 2.4.3, Fire Risk Evaluations and 2.4.4, Plant Change Evaluation of NFPA 805. These sections establish the requirements for the content and quality of the engineering evaluations to be used for approval of changes.*

The license conditions also defines the limitations imposed on the licensee during the transition phase of plant operations when the physical plant configuration does not fully match the configuration represented in the fire risk analysis. The limitations on self-approval are required because NFPA 805 requires that the risk analyses be based on the as-built, as-operated and maintained plant, and reflect the operating experience at the plant. Until the proposed implementation items and plant modifications are completed, the risk analysis is not based on the as-built, as-operated and maintained plant.

Overall, the revised license conditions provide structure and detailed criteria to allow self approval for FPP changes that meet the requirements of NFPA 805 with regard to Engineering Analyses, Fire Risk Evaluations and Plant Change Evaluations. The staff's evaluation of the Self-Approval Process for Fire Protection Program Changes (Post-Transition) is contained-in Section 2.6 otthis SE. The license conditions also reference the plant-specific modifications, and associated implementation schedules that must be accomplished at DAEC to complete

- 16 transition to NFPA 80S and comply with 10 CFR S0.48( c). In addition, the license conditions includes a requirement that appropriate compensatory measures will remain in place until implementation of the specified plant modifications is completed. These modifications and implementation schedules are identical to those identified elsewhere in the LAR, as discussed by the NRC staff in Sections 2.8.1 and 2.8.2, and explicitly reviewed in Section 3.0, of this SE.

Section 4.0 of this SE provides the NRC staff's review of the DAEC FPP license conditions.

2.4.3 Technical Specifications The NRC staff reviewed LAR Section S.2.2, "Technical Specifications" and Attachment 1\\1, "Technical Specification Changes," with regard to proposed changes to theDAEC TSs that are being revised or superseded during the NFPA 80S transition process. According to the LAR, the licensee conducted a review of the DAEC TSs to determine which, if any, TS sections will be impacted by the transition to a RI/PB FPP based on 10 CFR S0.48(c). The licensee identified changes to the Technical Specifications needed for DAEC adoption of the new fire protection licensing basis and provided applicable justification listed in Attachment N. The NRC staff found that the licensee had previously requested, and obtained NRC approval for, removal of fire protection requirements from the Duane Arnold TSs in Amendment 190 (Reference 40).

As a result of the licensee's removal of fire protection requirements from the Duane Arnold TSs, the NRC staff found that no additional changes to the Duane Arnold TSs'were required to*

support the NFPA 80S transition process.

2.4.4 Updated Final Safety Analysis Report (UFSAR)

The NRC staff reviewed the LAR and noted that Figure 4-8 of the LAR indicates that a revised UFSAR will be developed as a post-transition document representing the revised license conditions. The NRC staff noted that there was no implementation item to update the UFSAR, nor did the licensee provide any description of the changes that need to be made to the current UFSAR. In safe shutdown analysis (SSA) request for additional information (RAI) S, (Reference 41), the NRC staff requested that the licensee provide a description of the changes that will be made to the UFSAR as a result of the transition.

The licensee responded to SSA RAI S (Reference 12) and stated that the action to update the UFSAR was not included in Attachment S of the LAR since the revision of the UFSAR is required by regulation (10 CFR SO,71(e)) and part of the UFSAR update process. The licensee further stated that the need for an implementation item for NFPA 80S was not required and that the specific NEI 04-02 guidance reference is under revision in NRC FAa 12-0062, "U*FSAR Content", (Reference 39). The licensee further stated that the UFSAR.wili be revised to reflect transition to NFPA 80S in accordance with FAa 12-0062 and 10 CFR SO.71(e).

Since the licensee stated that the update to the UFSAR after approval of the LAR, will be in accordance with 10 CFRSO. 71 (e) and also that the content will be consistent with the guidance contained in NEI 04-02, the NRC staff concludes that the licensee's method to update the UFSAR following the guidance in FAa 12-0062 is acceptable.

2.5 Rescission of Exemptions

- 17 The NRC staff reviewed LAR Section 5.2.3, "Orders and Exemptions," Attachment 0, "Orders and Exemptions," and Attachment K, "Existing Licensing Action Transition," with regard to previously-approved exemptions to Appendix R to 10 CFR Part 50, which the transition to a FPP licensing basis in conformance with I\\IFPA 805 will supersede. These exemptions will no longer be required since upon approval of the RIIPB FPP in accordance with NFPA 805, Appendix R will not be part of the licensing basis for Duane Arnold.

The licensee requested and received NRC approval for 16 exemptions from 10 CFR Part 50 Appendix R These exemptions were discussed in detail in Attachment K of th~ lAR The NRC staff individually addressed the applicability and continuing validity of these exemptions as incorporated into the NFPA 805 FPP as part of the NRC staffs review of the appropriate section

.or fire area involved.

Disposition of Appendix R exemptions may follow two different paths during transition to NFPA805:

• The exemption was found to be unnecessary since the underlying condition has been evaluated using RI/PB methods (fire modeling and/or fire risk evaluation) and found to be acceptable and no further actions are necessary by the licensee.

• The exemption was found to be appropriate as a qualitative engineering evaluation that meets the deterministic requirements of NFPA 805 and is carried forward as part of the engineering analyses supporting NFPA 805 transition.

The following exemptions are rescinded as requested by the LAR and the underlying condition has been evaluated using RI/PB methods and found to be acceptable with no further actions (numbering scheme provided by the licensee):

~

Exemption 1- (19830426), Appendix R Exemption from Fire Protection Requirements of III.G.2 for Division 1 and Division 2 Cables Supplying the Scram Valves for Reactor Building North and South CRD Module Areas (1II.G.2 Criteria)

• Exemption 2 - (19830426), Appendix R Exemption from the Requirement to Provide Fixed Fire Suppression in the Control Room (1I1.G.3 Criteria)

Exemption 3 - (19831219), Appendix R Exemption for Fire Zone Boundaries Having Communication Paths with Less Than 3 Hour Fire Ratings Between Miscellaneous Doors and Dampers (11I.G.2.a Criteria)

Exemption 5 - (19831219), Appendix R Exemption from the Automatic Suppression Requirement for the Turbine Building Water Treatment and Condensate Pump Area (1II.G.2.c Criteria)

Exemption 6 - (19831219), Appendix R Exemption from the Requirement for Full Coverage by Automatic Suppression Systems in the [Heating, Ventilation, and Air Conditioning] HVAC Heat Exchanger and Chiller Area (lII.G.3 Criteria)

- 18 Exemption 7 - (19850701 ), Appendix R Exemption from the 8-Hour Battery Requirement for the Control Room (IILJ Criteria)

Exemption 8 - (19871014), Appendix R Exemption for Fire Zone Boundaries Having Communication Paths with Less Than 3 Hour Fire Ratings Between Zones (Doors No. 202 and 203) (1II.G.2.a Criteria)

Exemption 9 - (19871014), Appendix R Exemption from 3 Hour Rated Barrier in the Reactor Building Torus Area (1II.G.2.a Criteria)

Exemption 10 - (19871014), Appendix R Exemption from Automatic Suppression and Detection in the Reactor Building Torus Area (1II.G.2.b Criteria)

Exemption 11- (19871014), Appendix R Exemption from the Requirement for 3 Hour Fire Barriers in the Laydown Area and [Reactor Water Clean Up] RWCU Area (Fire Zone 3 Al3-B) (1II.G.2.a Criteria)

Exemption 12 - (19871014), Appendix R Exemption from the Requirement for 3 Hour Fire Barriers in the Reactor Building [residual heat removal) RHR Valve Room (Fire Zone 2-D) (1II.G.2.a Criteria)

Exemption 13 - (19871014), Appendix R Exemption from the Requirement for 3 Hour Rated Fire Barriers in the Equipment Hatch Between Fire Zones 3-B and 4-B (1II.G.2.a Criteria)

(I Exemption 14- (19871014), Appendix R Exemption from the Requirement of Separation of Redundant Trains of Safe Shutdown Cables and Equipment by 3 Hour Rated Fire Barriers for the Ventilation Duct Fire Dampers (1II.G.2.a Criteria)

The following exemptions are rescinded but the engineering evaluation of the underlying condition will be used as a qualitative engineering evaluation for transition to NFPA 805:

Exemption 4 - (19831219), Appendix R Exemption for Fire Zone Boundaries Having Communication Paths with Less than 3 Hour Fire Ratings Between* Zones (Equipment Hatch) (lILG.2.a Criteria)

Exemption 15 - (19871014), Appendix R Exemption from the Requirement that Structural Steel Forming Part of or Supporting Fire Barriers be Protected to a Fire Resistance Equivalent to that ofthe Barrier (1II.G.2.a Criteria)

Exemption 16 - (19910816),*Appendix R Exemption from the 3-Hour Fire Barrier Requirement for the Drywell Expansion Gap (1II.G.2.a Criteria) 2.6 Self-Approval Process for FPP Changes (Post-Transition)

Upon completion of the implementation of the RI/PB FPP and issuance of the license conditions discussed in Section 2.4.2 of this SE, changes to the approved FPP must be evaluated by the

- 19 licensee to ensure that they are acceptable. NFPA 805 Section 2.2.9, "Plant Change Evaluation," states the following:

In the event of a change to a previously approved fire protection program element, a RI plant change evaluation shall be performed and the results used as described in 2.4.4 to ensure that the public risk associated with fire-induced nuclear fuel damage accidents is low and that adequate defense-in-depth and safety margins are maintained.

NFPA 805, Section 2.4.4, "Plant Change Evaluation," states:

A plant change evaluation shall be performed to ensure that a change to a previously approved fire protection program element is acceptable. The evaluation process shall consist of an integrated assessment of the acceptability of risk, defense-in-depth, and safety margins.

2.6.1 Post-Implementation Plant Change Evaluation Process The NRC staff reviewed LAR Section 4.7.2, "Compliance with Configuration Control Requirements in Section 2.7.2 and 2.2.9 of NFPA 805," for compliance with the NFPA 805 plant change evaluation process requirements to address potential changes to the NFPA 805 RI/PB FPP after implementation is completed. The licensee developed a change process that is based on the guidance provided in NEI 04-02, Revision 2, (Reference 9), Section 5.3, "Plant Change Process," as well as, Appendices B, I, and J, as modified by RG 1.205, Revision 1, (Reference 8), Regulatory Positions 2.2.4, 3.1, 3.2, and 4.3.

LAR Section 4.7.2 states that the plant change process consists of four subtasks:

1. defining the change

2. preliminary risk screening

3. risk evaluation

4. acceptability determination In the LAR, the licensee stated that the plant change evaluation process begins by defining the change or altered condition in the LAR to be examined and the baseline configuration. The

. baseline is defined by the design basis and licensing basis. The licensee also stated that the baseline is defined as that plant condition or configuration that is consistent with the design basis and licensing basis and that the changed or altered condition or configuration that is not consistent with the design basis and licensing basis is defined as the proposed alternative.

The licensee stated that once the definition of the change is established, a screening will then be performed to identify and resolve minor changes to the FPP and that the screening will be consistent with fire protection regulatory review processes currently in place. The licensee further stated that the screening process will be modeled after NEI 02-03, "Guidance for Performing a Regulatory Review of Proposed Changes to the Approved Fire Protection Program," June 2003, (Reference 42), and that the process will address most administrative changes (e.g., changes to the combustible control program, organizational changes, etc.).

- 20 The licensee stated that once the screening process is completed, it will be followed by engineering evaluations that might include fire modeling and risk assessment techniques and that the results of these evaluations will then be compared to the acceptance criteria. The licensee further stated that changes that satisfy the acceptance criteria of NFPA 805 Section 2.4.4 and the fire protection license condition (see Attachment M to the LAR) can be implemented within the framework provided by NFPA 805, and that changes that do not satisfy the acceptance criteria cannot be implemented within this framework. The licensee further stated that the acceptance criteria will require that the resultant change in core damage frequency (CDF) and large early release frequency (LERF) be consistent with the fire protection license condition, and that the acceptance criteria will also include consideration of defense-in depth and safety margin, which would typically be qualitative in nature.

The licensee stated that the risk evaluation will involve the application of fire modeling analyses and risk assessment techniques to obtain a measure of the changes in risk associated with the proposed change and that, in certain circumstances, an initial evaluation in the development of the risk assessment may be a simplified analysis using bounding assumptions, provided the use of such assumptions does not unnecessarily challenge the acceptance criteria.

The licensee stated that the Plant Change Evaluations will be assessed for acceptability using the ~CDF (change in core damage frequency) and ~LERF (change in large early release frequency) criteria from the license con'ditions ~n9 that the proposed changes'will also be assessed to ensure it is consistent with the defense~in'-depth philosophy and that sufficient safety margins are maintained.

The licensee stated that its FPP configuration is defined by the program documentation and, to the greatest extent possible, the existing configuration control processes for modifications, calculations and analyses. The FPP license basis reviews will be used to maintain configuration control of the FPP documents. The licensee further stated that the configuration control procedures that govern the various DAEC documents and databases that currently exist will be revised to reflect the new NFPA 805 licensing bases requirements.

The licensee stated that several NFPA 805 document types such as: Nuclear Safety Capability Assessment (NSCA) supporting information, Non-Power Mode Review, Fire Modeling Reports, Fire Safety Assessments, risk evaluations, etc., generally require new control procedures and processes to be developed since they are new documents and databases created as a result of the transition to NFPA 805. In addition, the new procedures will be modeled after the existing processes for similar types of documents and databases. The licensee further stated that system level design basis documents will be revised to reflect the NFPA 805 role that the systems and components will play and that new procedures will be developed and existing documentation revised as part of license amendment implementation.

The process for capturing the impact of proposed changes to the plant on the FPP will continue to be a multiple step review and that the first step of the review will be an initial screening for process users to determine if there is a potential to impact the FPP as defined under NFPA 805 through a series of screening questions/checklists contained in one or more procedures depending upon the configuration control process being used, The licensee further stated that reviews'that identify potential FPP impacts will be sent to qualified individuals (e.g., Fire

- 21 Protection Engineer, Fire PRA Engineer, etc.) to ascertain the program impacts, if any, and that if FPP impacts are determined to exist as a result of the proposed change, the issue would be resolved by one of the following:

Deterministic Approach: Complying with NFPA 805 Chapter 3 and 4.2.3 requirements.

• PB Approach: Utilizing the NFPA 805 change process developed in accordance with NEI 04-02, RG 1.205, and the DAEC NFPA 805 fire protection license condition to assess the acceptability of the proposed change. This process will be used to determine if prior NRC approval of the proposed change is required.

The licensee stated that this process follows the requirements in NFPA 805 and the guidance

. outlined in RG 1.174, (Reference 25), which requires the use of qualified individuals, procedures that require calculations and evaluations be subject to independent review and verification, record retention,.peer review, and a corrective action program that ensures appropriate actions are taken when errors are discovered.

Based on the information provided by the licensee, the NRC staff concluded that the licensee's plant change evaluation process is considered acceptable because it meets the guidance in NEI 04-02, Revision 2, (~eference 9), as well as RG 1.205, Revision 1, (Reference 8), and addresses attributes for using Fire Risk Evaluations in accordance with NFPA 805.

Section 2.4.4 of NFPA 805 requires that Plant Change Evaluations consist of an integrated assessment of risk, defense-in-depth and safety margins. Section 2.4.3.1 of NFPA 805 requires

( that the probabilistic safety assessment (PSA) use CDF and LERF as measures for risk, Section 2.4.3.3 of NFPA 805 requires that the risk assessment approach, methods, and data shall be acceptable to the Authority Having Jurisdiction (AHJ) which is the NRC. Section 2.4.3.3 of NFPA 805 also requires that the PSA be appropriate for the nature and scope of the change being evaluated, be based on the as-built and as-operated and maintained plant, and reflect the operating experience at the plant.

The licensee's plant change evaluation process includes the required delta risk calculations, uses risk assessment methods acceptable to the NRC, uses appropriate risk acceptance criteria in determining acceptability, involves the use of a Fire PRA of acceptable quality, and includes an integrated assessment of risk, DID, and safety margins as discussed above.

2.6.2 Requirements for the Self Approval Process Regarding Plant Changes Risk assessments performed to evaluate plant change evaluations must use methods that are acceptable to the NRC staff. Acceptable methods to assess the risk of the proposed plant changes may include methods that have been used in developing the peer-reviewed Fire PRA model, methods that have been approved by the NRC via a plant-specific license amendment or through NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.

)

Based on the information provided by the licensee in the LAR, the process established to evaluate post-transition plant changes meets the guidance in NEI 04-02, Revision 2, (Reference 9), as well as RG 1.205, Revision 1, (Reference 8). The NRC staff concludes that the proposed plant change evaluation process at DAEC, which includes defining the change, a

- 22 preliminary risk screening, a risk evaluation, and an acceptability determination, as described in Section 2.6.1, is acceptable because it addresses the required delta risk calculations, uses risk assessment methods acceptable to the NRC, uses appropriate risk acceptance criteria in determining acceptability, involves the use of a Fire PRA of acceptable quality, and includes an integrated assessment of risk, DID, and safety margins.

However, before achieving full compliance with 10 CFR 50.48(c) by implementing the plant modifications listed in Section 2.7.1 of this SE (i.e., during full implementation of the transition to' NFPA 805), RI changes to the licensee's FPP may not be made without prior NRC review and approval unless the changes have been demonstrated to have no more than a minimal risk impact using the screening process discussed above because the risk analysis,is not consistent with the as-built, as-operated and maintained plant since the modifications have not been completed. In addition, the licensee is required to ensure that fire protection DID and safety margins are maintained during the transition process. The "Transition License Conditions" in the proposed NFPA 805 license condition include the appropriate acceptance criteria and other attributes to form an acceptable method for meeting Regulatory Position C.3.1 of RG 1.205, Revision 1, (Reference 8), with respect to the requirements for FPP changes during transition, and therefore demonstrate compliance with 10 CFR 50.48(c).

The proposed NFPA 805 license condition also includes a provision for self-approval of changes to the FPP that may be made on a qualitative, rather than RI basis. Specifically, the license conditions states that prior NRC review ar:J9 apprqyC)I';:ire not required for changes to the NFPA 805 Chapter 3 fundamental FPP elements"a'ncfdesign requirements for which an engineering evaluation demonstrates that the alternative to the NFPA 805 Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805 Chapter 3 element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement (i.e., has not

'impacted its contribution toward meeting the nuclear safety and radioactive release performance criteria), using a relevant technical requirement or standard.

Use of this approach does not fall under NFPA 805, Section 1.7, "Equivalency," because the condition can be shown to meet the NFPA 805 Chapter 3 requirement. Section 1.7 of NFPA 805 is a standard format used throughout NFPA standards. It is intended to allow '

owner/operators to use the latest state of the art fire protection features, systems, and equipment, provided the alternatives are of equal or superior quality, strength, fire resistance, durability, and safety. However, the intent is to require approval from the authority having jurisdiction because not all of these state of the art features are in current use or have relevant operating experience. This is a different situation than the use of functional equivalency since functional equivalency demonstrates that the condition meets the NFPA 805 code requirement.

Alternatively, the licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805 Chapter 3 elements are acceptable because the changes are "adequate for the hazard." Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805 Chapter 3 listed below, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the

- 23 change has not affected the functionality of the component, system, procedure, or physical arrangement (with respect to the ability to meet the nuclear safety and radioactive release performance criteria), using a relevant technical requirement o'r standard. NFPA 805 Section 2.4 states that engineering analysis is an acceptable means of evaluating a fire protection program against performance criteria. Engineering analyses shall be permitted to be qualitative or quantitative. Use of qualitative engineering analyses by a qualified fire protection engineer to determine that a change has not affected the functionality of the component, system, procedure or physical arrangement is allowed by NFPA 805 Section 2.4.

The four specific sections of NFPA 805 Chapter 3 for which prior NRC review and approval are not required to implement alternatives that an engineering evaluation has demonstrated are adequate for the hazard are:

1. "Fire Alarm and Detection Systems" (Section 3.8);

2. "Automatic and Manual Water-Based Fire Suppression Systems" (Section.3.9);

3. "Gaseous Fire Suppression Systems" (Section 3.10); and,

4. "Passive Fire Protection Features" (Section 3,11).

The engineering evaluations described above (i.e., functionally equivalent and adequate for the hazard) are engineering analyses governed by the NFPA 805 guidelines. In particular, this means that the evaluations must meet the requirements of NFPA 805, Section 2.4, "Engineering Analyses," and NFPA 805, Section 2.7, "Program Documentation, Configuration Control, and Quality." Specifically, the effectiveness of the fire protection features under review must be evaluated and found acceptable in relation to their ability to detect, control, suppress, and extinguish a fire and provide passive protection to achieve the performance criteria and not exceed the damage threshold for the plant being analyzed. The associated evaluations must also meet the documentation content (as outlined by NFPA 805, Section 2.7.1, "Content") and quality requirements (as outlined by NFPA 805, Section 2.7.3, "Quality") of the standard in order to be considered adequate. Note that the NRC staff's review of the licensee's compliance with NFPA 805, Sections 2.7.1 and 2.7.3 is provided in Section 3.8 of this SE.

According to the LAR, the licensee intends to use a Fire PRA to evaluate the risk of proposed future plant changes. Section 3.4.2, "Quality of the Fire Probabilistic Risk Assessment," of this SE discusses the technical adequacy of the Fire PRA, including the licensee's process to ensure that the Fire PRA remains current. The NRC staff determined that the quality of the licensee's Fire PRA and associated administrative controls and processes for maintaining the quality of the PRA model is sufficient to support self-approval of future RI changes to the FPP under the proposed license conditions, the NRC staff concludes that the licensee's process for self-approving future FPP changes is acceptable.

The NRC staff also concludes that the fire risk evaluation (FRE) methods used at DAEC to model the cause and effect relationship of associated changes as a means of assessing the risk of plant changes during transition to NFPA 805 may continue to be used after implementation of the RI/PB FPP, based on the licensee's administrative controls to ensure that the models remain current and to assure continued quality (see SE Section 3.4.2, "Quality of the Fire Probabilistic Risk Assessment"). Accordingly, these cause and effect relationship models may be used after transition to NFPA 805 as a part of the FREs conducted to determine the change in risk associated with proposed plant changes.

- 24 2.7*

Implementation Regulatory Position C.3.1 of RG 1.205, Revision 1, (Reference 8), says that a license condition included in a NFPA 805 LAR should include: (1) a list of modifications being made to bring the plant into compliance with 10 CFR 50.48(c); (2) a schedule detailing when these modifications will be completed; and (3) a statement that the licensee shall maintain appropriate compensatory measures in place until implementation of the modifications are completed.

2.7.1 Modifications The NRC staff reviewed LAR Attachment S, "Plant Modifications and Items to be Completed During Implementation," which describes the DAEC plant modifications necessary to implement the NFPA 805 licensing basis, as proposed. These modifications are identified in the LAR as necessary to bring DAEC into compliance with either the deterministic or PB requirements of NFPA 805. As described below, Attachment S provides a description of each of the proposed plant modifications, presents the problem statement explaining why the modification is needed, and identifies the compensatory actions required to be in place pending completion/implementation of the modification.

The NRC staff's review confirmed that the modifications identified in LAR Table S-1 are the

. same as those identified in LAR Table B-3, "Fire Area Transition," on a fire area basis, as the modifications being credited in the proposed NFPA 805 licensing basis. The NRC staff also confirmed that the LAR Table S-1 modifications and associated Table S-2 implementation schedule are the same as those provided in the proposed NFPA 805 license conditions.

LAR Table S-1 provides a detailed listing of the plant modifications that must be completed in order for DAEC to be in full compliance with NFPA 805 and implements many of the attributes upon which this SE is based and thereby meet the requirements of 10 CFR 50.48(c). These modifications will be implemented in accordance with the schedule provided in the proposed NFPA 805 license condition, which states that all modifications will be in place by December 31, 2014. In addition, the licensee has agreed to keep the appropriate compensatory measures in place until the modifications have been fully implemented.

2.7.2 Schedule LAR Section 5.4 provides the overall schedule for completing the NFPA 805 transition at DAEC.

The licensee stated that it will complete the implementation of the new program, including any procedure changes, process updates, and training to affected plant personnel to implement the NFPA 805 FPP within 180 days after NRC approval of the license amendment unless that falls within a scheduled outage window, which in that case, implementation would occur 60 days after startup from the scheduled outage.

LAR Section 5.4 also states that modifications will be completed by December 31, 2014 and that appropriate compensatory measures will be maintained until modifications are complete.

- 25 2.8 Summary of Implementation Items Implementation Items are items that the licensee has not fully completed or implemented as of the issuance date of the license amendment, but which will be completed during implementation of the license amendment to transition to NFPA 805 (e.g., procedure changes that are still in process, or NFPA 805 programs that have not been fully implemented). These items do not impact the bases for the safety conclusions made by the NRC staff in the associated SE. The licensee identified the implementation items in Attachment S, Table S-2 of the LAR. For each implementation item, the licensee and the NRC staff have reached a satisfactory resolution involving the level of detail and main attributes that each remaining change will incorporate upon completion.

Each implementation item will be completed prior to the deadline for implementation of the RIIPB FPP based on NFPA 805, as specified in the license condition and the letter transmitting the amended license (i.e., 180 days from the issuance date or 60 days after startup from a schedule outage).

The NRC staff, through an onsite audit or during a future fire protection inspection, may choose to examine the closure. of the implementation items, with the expectation that any variations discovered during this review, or concerns with regard to adequate completion of the implementation item, would be tracked and dispositioned appropriately under the licensee's corrective action program.

3.0 TECHNICAL EVALUATION

The following sections evaluate the technical aspects of the requested license amendment to transition the FPP at DAEC to one based on NFPA 805 (Reference 7) in accordance with 10 CFR50.48(c). While performing the technical evaluation of the licensee's submittal, the NRC staff used the guidance provided in NUREG 0800, Section 9.5.1.2, "Risk-Informed, Performance-Based Fire Protection," (Reference 28) to determine whether the licensee had provided sufficient information in both scope and level of detail to adequately demonstrate compliance with the requirements of NFPA 805, as well as the other associated regulations and guidance documents discussed in Section 2.0 of this SE. Specifically:

• Section 3.1 provides the results of the NRC staff review of the licensee's transition of the FPP from the existing deterministic guidance to that of NFPA 805 Chapter 3, "Fundamental FPP and Design Elements."

• Section 3.2 provides the results of the NRC staff review of the methods used by the licensee to demonstrate the ability to meet the nuclear safety performance criteria (NSPC).

• Section 3.3 provides the results of the NRC staff review of the fire modeling methods used by the licensee to demonstrate the ability to meet the NSPC using a fire modeling PB approach.

Section 3.4 provides the results of the NRC staff review of the fire risk assessments used to demonstrate the ability to meet the NSPC using a FRE PB approach.

- 26

• Section 3.5 provides the results of the NRC staff review of the licensee's Nuclear Safety Capability Assessment (NSCA) results by fire area.

Section 3.6 provides the results of the NRC staff review of the methods used by the licensee to demonstrate an ability to meet the radioactive release performance criteria.

• Section 3.7 provides the results of the NRC staff review of the NFPA 805 monitoring program developed as a part of the transition to a RIIPB FPP based on NFPA 805.

• Section 3.8 provides the results of the NRC staff review of the licensee's program documentation, configuration control and quality assurance.

In addition, Attachments A and B to this SE provide additional detailed information that was evaluated and/or dis positioned by the NRC staff to support the licensee's request to transition to a RIIPB FPP in accordance with NFPA 805 (Le., 10 CFR 50.48(c)). These attachments are discussed as appropriate in the associated section of this SE.

3.1 NFPA 805 Fu'ndamental FPP and Design Elements NFPA 805 (Reference 7) Chapter 3 contains the fundamental elements of the FPP and specifies the minimum design,requirements for fire protection systems and features that are necessary to meet the standard. The fundamental FPP elements and minimum design requirements include necessary attributes pertaining to the fire protection plan and procedures, the fire prevention program and design controls, injernal and external industrial fire brigades,

. and fire protection SSCs. However, 10 CFR 50.48(c) provides exceptions, modifications, and supplementations to certain aspects of NFPA 805, Chapter 3, as follows:

10 CFR 50.48(c)(2)(v) - Existing cables. In lieu of installing cables meeting flame propagation tests as required by Section 3.3.5.3 of NFPA 805, a flame-retardant coating may be applied to the electric cables, or an automatic fixed fire suppression system may be installed to provide an equivalent level of protection. In addition, the italicized exception to Section 3.3.5.3 of NFPA 805 is not endorsed. '

10 CFR 50.48(c)(2)(vi)

Water supply and distribution. The italicized exception to Section 3.6.4 of NFPA 805 is not endorsed. Licensees who wish to use the exception to Section 3.6.4 of NFPA 805 must submit a request for a license amendment in accordance with 10 CFR 50.48{c)(2)(vii).

10 CFR 50.48(c)(2){vii) - Performance-based methods. While Section 3.1 of NFPA 805 prohibits the use of PB methods to demonstrate compliance with the NFPA 805, Chapter 3 requirements, 10 CFR 50.48(c)(2)(vii) specifically permits that the FPP elements and minimum design requirements of NFPA 805, Chapter 3, may be subject to the PB methods permitted elsewhere in the standard.

. Furthermore, Section 3.1 of NFPA 805 specifically allows the use of alternatives to the NFPA 805, Chapter 3 fundamental FPP requirements that have been previously approved by

- 27 the NRC (which is the AHJ, as denoted in NFPA 805 and RG 1.205), and are contained in the currently approved FPP for the facility; 3.1.1 Compliance with NFPA 805, Chapter 3 Requirements The licensee used the systematic approach described in NEI 04-02, Revision 2 (Reference 9),

as endorsed by the NRC in Regulatory Guide 1.205, Revision 1 (Reference 8), to assess the proposed DAEC FPP against the NFPA 805, Chapter 3, requirements.

As part of this assessment, the licensee reviewed each section and subsection of NFPA 805, Chapter 3, against the existing DAEC FPP and provided 'specific compliance statements for each Chapter 3 attribute that contained applicable requirements. As discussed below, some subsections of NFPA 805, Chapter 3, do not contain requirements or are otherwise not applicable to DAEC, and others are provided with multiple compliance statements to fully document compliance with the element.

The methods used by DAEC for achieving compliance with the fundamental FPP elements and minimum design requirements are as follows:

1. The existing FPP element directly complies with the requirement: "noted in LAR Attachment A, "NEI 04-02 Table 8-1, Transition of Fundamental FP Program and Design Elements," (also called the 8-1 Table), as "G.omplies."

"-.~:

2. The existing FPP element complies through the use of an explanation or clarification:

noted in the LAR 8-1 Table as "Complies with clarification."

3. The existing FPP element complies through the use of existing engineering equivalency evaluations (EEEEs) whose bases remain valid and are of sufficient quality: noted in the LAR 8-1 Table as "Complies via Engineering Evaluation."

4. The existing FPP element complies with the requirement based on prior NRC approval of an alternative to the fundamental FPP attribute and the bases for the NRC approval remain valid: noted in the LAR 8-1 Table as "Complies by previous NRC approval."

5. The existing FPP element does not comply with the requirement, but the licensee IS requesting specific approval for a P8 method in accordance with 10 CFR 50.48(c)(2)(vii):

noted in the LAR 8-1 Table as "Submit for NRC Approval."

The NRC staff has determined that, taken together, these methods compose an acceptable approach for documenting compliance with the NFPA 805, Chapter 3 requirements, because the licensee has followed the compliance strategies identified in the endorsed NEI 04-02 guidance document.

The licensee stated in LAR Section 4.2.2, "Existing Engineering Equivalency Evaluation Transition," that it evaluated the EEEEs used to support compliance with the NFPA 805, Chapter 3, requirements in order to ensure continued appropriateness, quality, and applicability to the current DAEC plant configuration. The licensee determined that no EEEE used to support compliance with NFPA 805 required NRC approval.

- 28 EEEEs refer to "existing engineering equivalency evaluations" (previously known as Generic Letter 86-10 evaluations) performed for fire protection design variances such as fire protection

. system designs and fire barrier component deviations from the specific fire protection deterministic requirements. Once a licensee transitions to NFPA 805, future equivalency evaluations are to be conducted using a performance-based approach. The evaluation should demonstrate that the specific plant configuration meets the performance criteria in the standard.

Additionally, the licensee stated in LAR Section 4.2.3, "Licensing Action Transition," that the existing licensing actions used to demonstrate compliance have been evaluated to ensure that their bases remain valid. The results of these licensing action evaluations are provided in Attachment K of the LAR.

LAR Attachment A (the NEI 04-02 8-1 Table) provides further details regarding the licensee's compliance strategy for specific NFPA 805, Chapter 3 requirements, including references to where compliance is documented.

3.1.1.1 Compliance Strategy - Complies For certain NFPA 805, Chapter 3 requirements, as modified by 10 CFR 50.48(c)(2), the licensee determined that the RI/P8 FPP complies directly with the fundamental FPP element using the existing FPP element. In these instances, based on the validity of the licensee's statements, the NRC staff concludes that the licensee's statements of compliance are acceptable.

The following NFPA 805 sections, identified in LAR 8-1 Table, as complying via this method, and the applicable NFPA 805, Chapter 3 implementation items in LAR Attachment S, Table S-2 (Table S-2), required additional review by the NRC staff:

3.2.2.4

.3.2.3(3)

• 3.3.1.2(2)

.3.3.9

.3.4.1 (b).3.4.1 (c)

• 3.4.2.1 NFPA 805, Section 3.2.2.4, requires that the policy document identify the appropriate AHJ for the various areas of the FPP. The LAR 8-1 Table states, '.'Plant documentation will be u'pdated to include the statement that the NRC is the authority having jurisdiction (AHJ) for fire protection changes requiring approval." This update to plant documentation is addressed in Table S-2 as Implementation Item 1.

NFPA 805, Section 3.2.3(3), requires procedures to be established for performing reviews of the FPP, including related performance and trends. In LAR Table 8-1, the the licensee identified the current procedures for monitoring the FPP. The licensee also identified an implementation item as follows: "The monitoring program required by NFPA 805 will include a process that monitors and trends the FPP based on specific goals established to measure effectiveness."

This implementation item is included LAR Attachment S, Table S-2, Item 2.

I\\IFPA 805, Section 3.3.1.2(2), requires that plastic sheeting materials used in the power block be fire-retardant types that have passed NFPA 701, "Standard Methods of Fire Tests for Flame Propagation of Textiles and Films," large-scale tests or equivalent. In the LAR 8-1 Table, the licensee identified an implementation item to update plant documentation to include the

- 29 statement that "Plastic sheeting materials shall conform to the requirements of NFPA 701 or equivalent." This is addressed in Table S-2 as Implementation Item 3.

NFPA 805, Section 3.3.9, requires transformer oil collection basins and drain paths to be periodically inspected to ensure they are free of debris and functional, where these features are provided. In LAR Attachment A, for this attribute, the licensee stated the main, auxiliary, and startup transformers are provided with concrete dikes that drain to an oil collection pit. The licensee identified an Implementation Item to update procedures to include inspection of the oil collection basin and drain paths. This item is addressed in Table S-2 as Implementation Item 5.

NFPA 805, Section 3.4.1(b), requires that fire brigade members not have any other assigned plant duties that would prevent immediate response to fire or other emergencies. In LAR Attachment A, for this attribute, the licensee stated the fire brigade program documentation will be updated to clarify that brigade members will have no other assigned plant duties that would prevent immediate response. This item is addressed in Table S-2 as Implementation Item 7.

NFPA 805, Section 3.4.1 (c), requires that the fire brigade leader and at least two members have sufficient training and knowledge of nuclear safety systems to understand the effects of fire and fire suppressants on nuclear safety performance criteria. In LAR Attachment A, for this attribute, the licensee stated the fire brigade leader and only one other member are from the Operations Department. The licensee identified an implementation item to add a third member with nuclear safety systems knowledge and training or to assign a dedicated operations advisor.

This item is addressed in Table S-2 as Implementation Item 8. In an unnumbered Fire Protection Engineering (FPE) RAI (Reference 43), the NRC staff requested a description of how the requirements of this section of NFPA 805, specifically regarding sufficient training and knowledge of nuclear safety systems for the brigade leader and 'at least two brigade members, would be met. In its response (Reference 20), the licensee stated that the fire brigade leader is currently at least a Licensed Reactor Operator and at least two fire brigade members from the Maintenance or Radwaste Departments have completed training that includes the study of nuclear safety systems, their function, location and safety significance. In addition, the licensee concluded that since DAEC currently complies with the requirement and plans to maintain this level in the future, they retract Implementation Item 8. Based on this RAI response that DAEC will maintain fully qualified fire brigades, the NRC staff concludes that the licensee's statement of compliance is acceptable, because the licensee will maintain a fire brigade with a fire brigade leader that is qualified as a Licensed Reactor Operator and at least two additional fire brigade members that have training on nuclear safety systems. This type of qualification and training, J as described by the licensee, meets the NRC staff's expectation of sufficient knowledge and training as required by NFPA 805, Section 3.4.1(c).

NFPA 805, Section 3.4.2.1 requires that pre-fire plans detail the fire area configuration and fire hazards to be encountered, along with nuclear safety components and fire protection systems and features present. In LAR Attachment A, for this attribute, the licensee identified an implementation item to update the pre-fire plans to reflect the RIIPB program. The update includes components necessary to achieve the NSPC; which require entry to the fire affected area as well as equipment and portions of the fire affected area where RI/PB analysis rely on assumptions that could be impacted by fire brigade performance. This item is addressed in Table S-2 as Implementation Item 9.

- 30 Based on the licensees statement of compliance and the associated Implementation Items as described in LAR Attachment A and listed in LAR Attachment S for the individual attributes described above, as well as the statements that these items will be complete prior to implementation; the NRC staff concludes the licensee's statements of compliance are acceptable because completion of the implementation items will bring these attributes into compliance with the requirements.

3.1.1.2 Compliance Strategy - Complies with Clarification For certain NFPA 805, Chapter 3 requirements, the licensee provided additional clarification when describing its means of compliance with the fundamental FPP element. In these instances, the NRC staff reviewed the additional clarifications and concludes that the licensee will meet the underlying requirement for the FPP element as clarified.

The following NFPA 805 sections identified in the LAR B-1 Table as complying via this method required additional review by the NRC staff:

3.2.3(1) 3.5.16 NFPA 805, Section 3.2.3(1) provides the requirement for establishing procedures that address inspection, testing, and maintenance for fire protection systems and features credited by the' FPP. In the LAR B-.1 Table, the licensee stated in the compliance basis that surveillance frequencies may be modified in accordance with the methodology in EPRI Technical Report No.

TR1006756, "Fire Protection Equipment Surveillance Optimization and Maintenance Guide,"

(Reference 44). EPRI TR1006756 is published by the Electric Power Research Institute and provides guidance for licensees follow in order to optimize their fire protection surveillance and testing practices and frequencies for fire protection systems, structures, and components based upon performance. In FPE RAI 2, (Reference 41), the NRC staff informed the licensee that the proposed future use of TR 1006756 methods was not a clarification to compliance and requested information from the licensee regarding the planned application and incorporation of these methodologies at DAEC. In its response to the RAI (Reference 12), the licensee Indicated that the appropriate FPP document(s) will be updated to provide a requirement that if the plant elects to implement the methodologies in EPRI Report TR1006756, that the methodologies will be implemented in their entirety as they pertain to the fire protection systems or features being evaluated. In addition, the licensee added Implementation Item 18 to LAR Attachment S, Table S-2, to update the appropriate FPP documents.

Based on the NRC staffs review, the surveillance methods in TR1006756 are considered to be PB alternatives to NFPA 805, Chapter 3, and as such, require an approval by the NRC in accordance with the requirements of 10 CFR 50,48(c)(2)(vii). In FPE RAI12 (Reference 45),

the NRC staff requested the licensee submit a request for approval for the use the EPRI methods at DAEC, in accordance with 10 CFR 50,48(c)(2)(vii). In its response to FPE RAI 12 (Reference 16), the licensee provided an amendment to the LAR, Attachment L, to include a new Approval Request which requests NRC approval for use of the TR1 006756 methods in accordance with 10 CFR 50,48( c)(2)(vii). The NRC staff review of this item is documented in SE Section 3.1,4.1.

NFPA 805, Section 3.5.16 requires that the fire protection water supply be dedicated for fire protection use only, subject to Exceptions 1 and 2 described in the standard. Exception 1

- 31 allows fire water supply to provide backup to nuclear safety systems provided the combined fire protection and nuclear safety flow demands can be met. Exception 2 allows fire protection water storage to be provided by plant systems serving other functions provided the storage has a dedicated capacity capable of providing the maximum fire protection demand.

The LAR B-1 Table for this attribute states that DAEC complies with both Exceptions 1 and 2 of Section 3.5.1*6, and provides clarification on the combined use of the fire protection system and water supply. In FPE RAI 8 (Reference 41), the NRC staff requested the licensee provide additional information on the availability of both fire pumps for all areas where a fire pump is credited to support the nuclear safety function, and additional discussion on how sufficient water volume is provided to supply both the fire demand and the demand of other plant systems when a fire pump is used to supply these systems. In its response to the RAI (Reference 12), the licensee stated that upon further evaluation, DAEC does not require compliance with Exception 1 because the fire protection water system is not credited in the NSCA or the Fire PRA.

With regard to Exception 2, the licensee stated that the wet pit volume is sufficient to meet fire water demands and fire pumps are not used to supply any other plant systems for nuclear safety. Makeup water to the pit is provided from river water pumps that are protected such that at least two pumps are available for a fire in any fire area. On the basis that the fire water system is not credited with providing water to any othe*r plant systems for nuclear safety, the wet pit has adequate volume for the largest sprinkler and hose stream demand, and makeup capability to the wet pit is protected, the NRC staff concludes that the licensee's statement of compliance with NFPA 805, Section 3.5.16 is acceptable because it meets the intent of this element, which is to ensure that the dedicated fire water system can supply sufficient water to satisfy fire protection system demands.

In review of the water supply and pump configuration for DAEC, the NRC staff identified that circulating water pumps and fire pumps take suction from a common wet pit. This configuration could be.a concern for a Turbine Building (TB) fire that results in failu~e of a circulating water system connection at the condenser and the subsequent loss of wet-pit volume through the failed connection. If not addressed, the circulating water pumps could rapidly deplete the wet pit and cause a loss of suction, and subsequent failure of the fire pumps. In FPE RAI 11 (Reference 45), the NRC staff requested that the licensee describe the recovery actions (RAs) and procedures that ensure that sufficient volume is maintained in the wet-pit under this scenario to provide adequate suction to the fire pumps. In its response (Reference 16), the licensee stated that procedures are in place to direct operators to trouble shoot and restore water to the wet pit. Loss of water would be identified by alarms or trip of the system. Based on the licensee's statements that low water level will annunciate in the control room, circulating water pumps trip on low level in the wet pit, and procedures are in place to restore the water supply, the NRC staff concludes that the licensee response is acceptable because it meets the intent of NFPA 805, Section 3.5.16, which is to ensure that the dedicated fire water system can supply sufficient water to satisfy fire protection system demands.

- 32 3.1.1.3 Compliance Strategy - Complies with Use of EEEEs For certain NFPA 805, Chapter 3 requirements, the licensee demonstrated compliance with the fundamental FPP element through the use of EEEEs. For NFPA 805 Sections 3.3.1.3.1, 3.4.1 (a)(1), and 3.8.1, the licensee has identified/Implementation Items 4, 6, and 10 in Table S-2, respectively, to address additional NFPA code requirements. The NRC staff reviewed the licensee's statement of continued validity for the EEEEs, the statement on the quality and appropriateness of the evaluations, and the identified implementation items and concludes that the licensee's statements of compliance in these instances are acceptable.

NFPA 805 section 3.3.8 was identified in LAR B-1 Table as complying via the use of EEEEs required additional review by the NRC staff.

NFPA 805 Section 3.3.8 provides requirements for bulk*storage of flammable and combustible liquids. The compliance statement for this attribute does not address the portion of the requirement that prohibits bulk storage inside structures containing SSCs important to nuclear safety. In FPE RAI 5 (Reference 41), the NRC staff requested that the licensee describe how DAEC complies with the prohibition in Section 3.3.8 for bulk storage of flammable and combustible liquids in structures containing SSCs important to nuclear safety. In its response*

(Reference 12), the licensee stated that DAEC complies with Section 3.. 3.8 and there is no bulk storage in structures containing SSCs important to nuclear safety. Flammable and combustible liquids in vessels that are installed as part of a design system (e.g., day tanks for diesel generators, fire pumps, or turbine lube oil) are not considered subject to this prohibition. The NRC staff concludes that this response is acceptable because it describes compliance with the requirement.

3.1.1.4 Compliance Strategy, - Complies with Previous NRC Approval Certain NFPA 805, Chapter 3 requirements were supplanted by an alternative that was previously approved by the NRC. The approval was documented in the original 1978 FPP SER (Reference 46) and a 1980 NRC letter on the status of SER open items (Reference 47).

In each instance, the licensee evaluated the basis for the original NRC approval and determined that in all cases the bases were still valid. The NRC staff reviewed the information provided by the licensee and concludes that previous NRC approval had been demonstrated using suitable documentation that meets the approved guidance contained in RG 1.205, Revision 1. Based on the licensee's justification for the continued validity of the previously approved alternatives to the NFPA 805, Chapter 3, requirements, the NRC staff concludes that the licensee's statements of compliance in these instances are acceptable.

3.1.1.5 Compliance Strategy - Submit for NRC Approval The licensee also requested approval for the use of PB methods to demonstrate compliance with fundamental FPP elements. In accordance with 10 CFR 50.48(c)(2)(vii), the licensee requested specific approvals be included in the license amendment approving transition to NFPA 805 at DAEC. The NFPA 805 sections identified in LAR B-1 Table as complying via the PB method are as follows:

- 33 3.2.3(1), which concerns the establishing of procedures for inspection, testing, and maintenance for fire protection systems and features. The licensee requests the use of EPRI Report TR1006756, "Fire Protection Equipment Surveillance Optimization and

. Maintenance Guide," to modify fire protection system surveillance frequencies. This approval request was added by the licensee in response to FPE RAI 12. See SE Section 3.1.1.2 for further information regarding related RAls. See Section 3.1.4.1 of this SE for the NRC staff's safety evaluation on this request.

3.3.3, which concerns the classification of interior floor finish in accordance with NFPA 101 (Reference 48), Class I criteria. The licensee requests the use of epoxy floor coatings that do not meet the exact interior finish requirements of NFPA 805. See

$ection 3.1.4.2 of this SE for the NRC staff's safety evaluation on this request.

• 3.3.5.2, which concerns the use of metal tray and metal conduit for electrical raceways.

The licensee requests approval for the use of plastic embedded conduit installations.

See Section 3.1.4.3 of this SE for the NRC staff's safety evaluation on this request.

3.5.11, which concerns the design configuration of sectionalizing valves for isolation of fixed and manual suppression systems. The licensee requests approval for installed systems where closure of sectioning valves could isolate both fixed suppression systems and hose stations provided for manual backup. See Section 3.1.4.4 of this SE for the NRC staff's safety evaluation on this request.

As discussed in SE Section 3.1.4 below, the NRC staff concluded that the use of PB methods to demonstrate compliance with these fundamental FPP elements is acceptable.

3.1.1.6 Compliance Strategy - Multiple Strategies In certain compliance statements of the NFPA 805, Chapter 3 requirements, the licensee used more than one of the above strategies to demonstrate compliance *with aspects of the fundamental FPP element.

In each of these cases, the NRC staff concludes that the individual compliance statements are acceptable, for the reasons outlined above, that the combination of compliance strategies is acceptable, and that holistic compliance with the fundamental FPP element is assured.

3.1.1.7 Chapter 3 Sections Not Reviewed Some NFPA 805 Chapter 3 sections either do not apply to the transition to a RI/PB FPP or have no technical requirements. Accordingly, the NRC staff did not review these sections for acceptability. The sections that were not reviewed fall into one of the following categories:

Sections that do not contain any technical requirements. (e.g., NFPA 805 Chapter 3, Section 3.4.5 and Section 3.11).

Sections that are not applicable to DAEC because of the following:

- 34 The licensee stated that DAEC does not have systems of this type installed (e.g" Section 3:6.5, which applies to seismic hose station cross-connected to non fire protection systems and Section 3.10A, which has single failure limitations for gaseous fire suppression systems in areas that are required by both. primary and backup systems); and The requirements are structured with an applicability statement (e.g., Section 3.3.12, which applies to reactor coolant pumps in non-inerted containments, or Sections 3A.1 (a)(2) and 3A.1 (a)(3), which code(s) apply to the fire brigade depends on the type of brigade specified in the FPP at the site).

3.1.1.B Compliance with Chapter 3 Requirements Conclusion As discussed above, the !\\IRC staff evaluated the results of the licensee's assessment of the proposed DAEC RIIPB FPP against the NFPA B05, Chapter 3, fundamental FPP elements and minimum deSign requirements, as modified by the exceptions, modifications, and supplementations in 10 CFR 50AB(c)(2). Based on this review of the licensee's submittal. as

. supplemented, the NRC staff concludes that the RI/PB FPP is acceptable with respect to the fundamental FPP elements and minimum design requirements of NFPA B05, Chapter 3, as modified by 10 CFR 50AB(c)(2), because the licensee accomplished the following:

Used an overall process consistent with NRC staff approved guidance to determine the state of compliance w,ith each of the applicable NFPA 805, Chapter 3 requirements.

III Provided appropriate documentation of DAEC's state of compliance with the NFPA B05 requirements, which adequately demonstrated compliance in that the licensee was able to substantiate that it complied:

With the requirement directly or with the requirement directly after the completion of an implementation item.

With the intent of the requirement (or element) given adequate justification.

Via previous NRC staff approval of an alternative to the requirement Through the use of EEEEs.

Through the use of a combination of the above methods.

Through the use of a PB method that the NRC staff has specifically reviewed and approved in accordance with 10 CFR 50AB(c)(2)(vii).

3.1.2 Identification of the Power Block The NRC staff reviewed the DAEC structures identified in LAR Table 1-1 "Definition of Power Block" as composing the "power block." The plant structures listed are established as part of the power block for the purpose of denoting the structures and equipment included in the DAEC RI/PB FPP that have additional requirements in accordance with 10 CFR 50AB(c) and

- 35 NFPA 805. As stated in the LAR, Section 4.1.3, the power block includes structures that contain equipment that could affect plant operation for power generation; equipment important to safety; equipment that could affect the ability to maintain NSCA in the event of a fire; or.

structures containing radioactive materials that could potentially be released in the event of a fire. The NRC staff concludes that the licensee has appropriately evaluated the structures al')d equipment at DAEC, and adequately documented a list of those structures that fall under the definition of "power block" in NFPA 805.

3.1.3 Closure of Generic Letter 2006-03, "Potentially Nonconforming Hemyc' and MTTM Fire Barrier Configurations," Issues DAEC does not use either the Hemyc'or MTTM electrical raceway fire barrier systems (ERFBS). Therefore, the generic issue (GL 2006 Reference 49) related to the use of these ERFBS is not applicable to DAEC. GL 2006-03 requested that licensees evaluate their facilities to confirm compliance with existing applicable regulatory requirements in light of the results of NRC testing that determined that both Hemyc and MT fire barriers failed to provide the protective function intended for compliance with existing regulations, for the configurations

. tested using the NRC's thermal acceptance criteria.

3.1.4 Performance Based Methods for NFPA 805, Chapter 3, Elements In accordance with 10 CFR 50.48(c)(2)(vii), a licensee may request NRC approval for use of the PB methods permitted elsewhere in the standard as a means of demonstrating compliance with the prescriptive NFPA 805, Chapter 3, Fundamental FPP Elements and Minimum Design Requirements. Paragraph 50.48(c)(2)(vii) of 10 CFR requires that an acceptable PB approach

. accomplish the following:

(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 safe shutdown capability).

In Attachment L, "NFPA 805, Chapter 3, Requirements for Approval 10 CFR 50.48(c)(2)(vii)," of the LAR, the licensee requested NRC staff review and approval of PB methods to demonstrate an equivalent level of fire protection for the requirement of the elements identified in Section 3.1.1.5 of this SE. In addition, in response to FPE RAI 12 (Reference 16), the licensee supplemented Attachment L to include an additional request related to NFPA 805 Section 3.2.3(1) as described in SE Section 3.1.1.2. The NRC staff evaluation of these proposed methods is provided below.

3.1.4.1 NFPA 805, Section 3.2.3(1) -Inspection, Testing, and Maintenance Procedures As discussed in SE Section 3.1.1.2, the NRC staff requested information regarding the compliance strategy for NFPA 805 Section 3.2.3(1). In its response to FPE RAI 12

- 36 (Reference 16), the licensee supplemented LAR Attachment L with an additional Approval Request. The licensee requested NRC staff review and approval of a PB method to demonstrate an equivalent level of fire protection for the requirements regarding inspection, testing, and maintenance of credited fire protection systems and features. Specifically, the licensee requested approval to use PB methods to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features.

As described by the licensee, the PB inspection, testing, and maintenance frequencies would be

'established using the methods described in EPRI Technical Report TR-100675~, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features", Final Report, July 2003 (Reference 44).

The licensee stated that there will be no impact on the NFPA 805 NSCA because the use of PB test frequencies established per TR1006756 methods, combined with NFPA 805 Section 2.6, "Monitoring Program," will provide assurance that the availability and reliability of the fire protection systems and features are maintained to the levels assumed in the NFPA 805 engineering analyses and ensure that there is no impact on the systems and features ability to perform its function.

The radiological release performance criteria are satisfied based on the determination of limiting radioactive release (LAR Attachment E). Some fire protection systems and features are credited as part of that evaluation and the use of PB test frequencies established per TR1006756 methods, with the new Monitoring Program, should ensure that the availability and reliability of the systems and features are maintained to the levels assumed in the analyses credited for meeting the Radioactive Release performance criteria. Therefore, there should be no adverse impact on meeting these criteria.

The licensee further stated that the proposed alternative maintains the safety margins of the analyses because it provides assurance that the availability and reliability of the systems and features are maintained to the levels assumed in the original NFPA 805 engineering analyses which includes those assumptions credited in the FRE safety margin discussions. In addition, the licensee stated the use of these methods should in no way invalidate the inherent safety margins contained in the NFPA codes used for design and maintenance of fire protection systems and features. Therefore, the safety margin inherent and credited in the analyses should be preserved.

The three echelons of DID 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, RAs).

The TR1006756 PB methods will be used to establish PB inspection, testing, and maintenance frequencies for fire protection systems and features credited by the FPP. Therefore, the TR 1006756 PB methods do not affect echelon 1. For echelons 2 and 3, the use of PB test frequencies methods in TR1006756, combined with the new monitoring program, should provide assurance that the availability and reliability of the fire protection systems and features

- 37 credited for DID are maintained to the levels assumed in the NFPA 805 engineering analyses.

Therefore, there should be no adverse impact to DID echelons 2 and 3.

Based on its review of the information submitted by the licensee, and in accordance with 10 CFR 50.48(c)(2)(vii), the NRC staff concludes that the proposed PB method is an acceptable alternative to the corresponding NFPA 805, Section 3.2.3(1) requirement because it satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release, maintains sufficient safety margin, and maintains adequate fire protection DID.

3.1.4.2 NFPA 805, Section 3.3.3 - Interior Finish The licensee requested review and approval of a PB method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.3.3 regarding interior finishes.

Specifically, the use of an epoxy floor coating that does not meet the specific combustibility standards for interior finish cited in Section 3.3.3.

As described by the licensee, this PB method would consist of reliance on previous NRC resolution of the use of epoxy floor coatings at DAEC, including the supporting testing and analysis. The DAEC epoxy floor covering was the subject of an Unresolved Issue (URI) from an NRC inspection in 2003 (50-331/03-02) (Reference 50) and NRC InformationNotice IN 2007-26, "Combustibility of Epoxy Floor Coatings at Commercial Nuclear Power Plants" (Reference 51). The licensee response to the URI (Reference 52), that provided an evaluation of the coating and the potential to c'ontribute to the fire area combustible loading, impact on areas required to be free of combustibles (e.g., separation zones), and fire propagation from one fire area to another. The evaluation determined that epoxy floor coatings were not a significant contributor to combustibles, or fire propagation between separated components or fire areas. On the basis of the licensee's evaluation, the NRC staff closed the URI in NRC Inspection Report (50-331/05-09) (Reference 53).

The licensee stated that the use of this alternative for interior finish would have no adverse,

impact on combustibility or fire propagation considerations associated with floor finishes. The licensee concluded that the combustible loading in safety related areas and challenges to plant fire barriers will not be adversely impacted by the proposed change because:

• The floor coating thickness meets the criteria in NFPA 805 for limited combustible material with only a few exceptions;

• DAEC's determination that the contribution of the floor coating to combustible loads is negligible;

• The coatings permitted at DAEC are either NFPA Class A or American Society for Testing and Material (ASTM) E84 (Reference 54) tested with a flame spread less than 50; and

• The epoxy coating is on the *noar.

- 38 The ASTM E84 test is conducted with the material on the ceiling of a tunnel. This configuration would allow the flame to directly impinge on the ceiling surface, enhancing flame spread. With the material on the floor, the heat flux t6 the surface is much less than would be expected in the ceiling configuration since the convective flame is directing the heat away from the surface.

Therefore, the overall flame spread should be less, even with a slightly greater thickness.

During the review of this request, the NRC staffidentified that the discussion in the request only addressed the negligible impact of the floor coating with regard to challenging the plant's fire barriers, but did not specifically address the issue of propagation in separation areas required to be free of combustibles. In FPE RAI 3 (Reference 41), the NRC staff asked the licensee to clarify that the epoxy coatings will not propagate fire between spatially separated nuclear safety capability systems and components. In its response (Reference 12), the licensee provided additional clarification that the NRC staff's closure of the original URI was based on the evaluation of both the potential for fire spread in areas crediting spatial separation with no intervening combustibles, and for fire spread between fire areas past a single door providing separation. Based on review of the cited references and the licensee's statements in the RAI, the NRC staff concludes that this clarification is acceptable because there is reasonable assurance that a fire involving this epoxy floor coating will not propagate across credited spatial separations nor pastdoors credited for fire separation.

In the request, the licensee stated that there will be no impact on the NFPA 805 nuclear safety performance goals, performance objectives, and performance criteria because the use of epoxy floor coating does not affect nuclear safety as it, in general, meets the definition of a limited combustible material with isolated thickness excesses. The licensee stated the floor coating materials were evaluated to have a negligible effect on combustibility, and application of epoxy coatings is controlled via a DAEC procedure to ensure that the amount of material does not add appreciable amounts of combustible material to the plant. The licensee also stated that this alternative will have no effect on the NFPA 805 r9diological release performance criteria, since the radiological release review was performed based on the manual fire suppression activities in areas containing or potentially containing radioactive materials and is not dependent on the floor coating materials. The coatings do not change the radiological release evaluation, which concluded that potentially contaminated water is contained and smoke monitored. Floor coatings do not add additional radiological materials to an area or challenge system boundaries.

The licensee further stated in the LAR that the proposed alternative maintains the safety margins of the licensee's analyses related to combustible loading, because the proposed alternative will not alter the methods, input parameters, and acceptance criteria used in evaluating fire barrier performance or fire related damage to plant systems and equipment as related to combustible loading. In FPE RAI 4 (Reference 41), the NRC staff asked for clarification on the safety margin discussion in support of the request. In its response (Reference 12), the licensee stated that fhe use of epoxy floor coating does not affect safety margins as it, in general, meets the definition of a limited combustible material and the coatings were evaluated to have negligible effect on combustibility. The application of floor coatings is controlled by procedure and the precautions and limitations on the use of these materials do not impact the analysis of the fire event. Based on the licensee's statements in the RAI response, the NRC staff concludes that the clarification of the margin of safety statements is acceptable because it meets the intent of NFPA 805, Section 3.3.3, which is to ensure that the interior

- 39 finishes, including floor coatings, have design controls to restrict the use of combustible materials.

Finally, the licensee stated in the LAR that fire protection DID will be maintained, because the use of epoxy floor coatings does not affect echelons 1, 2 and 3. The use of epoxy floor coatings

'is not a fire prevention feature and does not compromise fire detection, suppression or control, or impact post-fire safe shutdown capability.

Based on its review of the information submitted by the licensee, and in accordance with 10 CFR 50.48(c)(2)(vii), the NRC staff concludes that the proposed PB method is an acceptable alternative to the corresponding NFPA 805, Section 3.3.3 requirement because it satisfies the performance goals, objectives, and criteria specified in NFPA 805 related to nuclear safety and radiological release, maintains sufficient safety margin, and maintains adequate DID.

3.1.4.3 NFPA 805, Section 3.3.5.2 - Metal Tray and Metal Conduit for Electrical Raceways The licensee requested NRC staff review and approval of a PB method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805 Section 3.3.5.2 regarding use of metal tray or conduit for electrical raceways. Specifically, the licensee has requested approval of a PB method to justify the use of plastic conduit embedded in concrete.

As described in the request, plastic conduit is used in embedded installations; Access points to the conduit are required to be rigid steel. The plastic conduit is embedded in concrete and protected from mechanical damage, exposure fire, and fire within the conduit will not expose external targets.. The licensee requested approval on the following basis:

• The plastic conduit, while a combustible material, is not subject to flame or heat impingement froll) an external source, which would result in structural failure, contribution to fire load, or damage to the circuits contained within where the conduit is embedded in concrete; and Failure of the circuits within the conduit resulting in fire would not result in damage to external targets.

The licensee stated that the use of embedded conduit in locations does not affect nuclear safety as the material in which conduits are run are not subject to the failure mechanisms potentially resultant in internal circuit damage; or damage to and from external targets. Therefore there should be no impact on the NSCA. The licensee also stated that th,e use of embedded plastic conduit should have no effect on the NFPA 805 radiological release performance criteria, because the embedded installations have no impact on radiological release. The radiological release review was performed based on the manual fire suppression activities in areas containing or potentially containing radioactive materials and is not dependent on the type of conduit material used.

The licensee further stated that, 'The plastic conduit material is embedded in a non-combustible configuration. The use of these materials has been defined by the limitations of the analytical methods used in the development of the Fire PRA. Therefore, the inherent safety margin and

- 40 conservatisms in these methods remain unchanged.'~ In FPE RAI 4 (Reference 41), the NRC staff asked the licensee to clarify the safety margin discussion in support of this request. In its response (Reference 12), the licensee stated the plastic material is embedded in non combustible configurations. The material is protected from mechanical damage and from damage resulting from either an exposure fire or a fire from within the conduit impacting other targets. The precautions and limitations on the use of embedded conduit do not impact the fire analysis and therefore, inherent safety margins are maintained. Based on the limitation of plastic conduit to embedded locations as described in the LAR and the licensee's statements in the RAI response, the NRC staff concludes that the clarification of the margin of safety statements is acceptable because it meets the intent of NFPA 805, 'Section 3.3.5.2, which is to ensure that construction and materials used in electrical raceways provide adequate protection for the cables inside from a fire outside the raceway.

Finally, the licensee stated that fire protection DID will be maintained, because the use of plastic embedded conduit in concrete does not affect echelons 1, 2 and 3. The conduit is not a fire prevention feature and does not compromise fire detection, suppression or control, or impact post-fire safe shutdown capability. The NRC staff concludes that this DID evaluation is acceptable because the use of plastic embedded conduit in concrete does not adversely affect any of the DID echelons.

Based on its review of'the LAR, as supplemented, and in accordance with 10 CFR 50.48(c)(2)(vii), the NRC staff concludes that the proposed PB method is an acceptable alternative to the corresponding NFPA 805, Section 3.3.5.2 requirement because it satisfies the performance goals, objectives, and criteria specified in NFPA 805 related to nuclear safety and radiological release, maintains sufficient safety margin, and maintains adequate fire protection DID.

3.1.4.4 NFPA 805, Section 3.5.11 - Common Isolation of Fire Water Supply to Fixed Systems and Backup Manual Hose Stations The licensee requested NRC staff review and approval a PB method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805 Section 3.5.11 regarding' means to isolate the yard fire main loop without simultaneously isolating fixed suppression systems and backup manual hose stations. Specifically, the request was for approval of a PB method to justify plant design configurations that do not meet NFPA 805 Section 3.5.11.

As described in LAR Attachment L, Approval Request 3, DAEC identified two locations where the water supply to both fixed fire suppression systems and fire hose stations provided for manual backup may need to be isolated as a result of maintenance or repair and therefore do not meet the exact requirements of NFPA 805, Section 3.5.11. These locations are in the Pump House and the Control Room Heating Ventilation and Air Conditioning (HVAC) Room. The licensee requested approval for the fire protection system water supply isolation configuration on the basis of:

Backup suppression is readily available from alternative sources;

~ 41 The fire brigade is trained and has access to hose lines connected to the unaffected yard fire water system which provides backup fire suppression in the event of a loss of suppression system and manual hose station water; and

• The control room HVAC room is protected by smoke detection, charcoal filter bed thermal detection, and an area sprinkler system supplied by an independent water supply.

For the pump house, Sprinkler System 7, which protects the diesel fire pump and day tank rooms, is supplied from the same header as the pump house standpipe. For the Control Room HVAC room, manual Deluge Systems 21 and 22, which protect the charcoal beds in the Control Building Standby Filter units, are supplied from the TB standpipe system, which also supplies the hose stations in this area. The LAR stated that an alternative manual suppression capability is available from yard hydrants and the HVAC room has smoke detection and an automatic sprinkler system (Sprinkler System 12). Thermal detectors are also provided monitor the internal charcoal beds temperature.

In FPE RAI 7 (Reference 41), the NRC staff requested additional discussion regarding the specific hose stations impacted by the common supply issue, the actions necessary to provide backup suppression, including the distance from the yard hydrants to the affected areas, and clarification regarding additional systems in the pump house that appeared to be affected by a common isolation point. In its response (Reference 12), the licensee identified the impacted hose stations, described the distance and capability to provide suppression from yard hydrants, and added Sprinkler System 21 in the pump house that is also impacted by isolation of the pump house header.

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The licensee stated that the common isolation of fixed systems and manual backup hose stations does not affect nuclear safety as the configuration of the Control Building Standby Filter Unit deluge systems to the TB standpipe water supply and the diesel fire pump and day tank room suppression system to the Pump house standpipe water supply does not affect nuclear safety. There are alternative measures available to ensure suppression of a fire if one were to occur in these areas. Backup fire suppression for these areas is manual suppression by the fire brigade using an alternative water supply. The alternative water supply would be a hose connection to the main fire water system via yard fire hydrants. Therefore, there is no impact on the NSCA. The licensee also stated that the configuration of the Control Building Standby Filter Unit deluge systems to the TB standpipe water supply and the diesel fire pump and day tank room suppression system to the Pump house standpipe water supply has no impact on the radiological release performance criteria. The radiological release review was performed based on the manual fire suppression activities in areas containing or potentially containing radioactive materials and is not dependent on the suppression system water supplies. The suppression system water supplies do not change the radiological release evaluation, which concluded that*

potentially contaminated water is contained on site and smoke monitored.

The licensee further stated that, 'The configuration of the Control Building Standby Filter Unit deluge systems to the Turbine Building standpipe water supply system and the diesel fire pump and day tank room suppression system to the Pump house standpipe water supply system does not affect safety margin. The use of these systems has been defined by the limitations of the analytical methods used in the development of the Fire PRA. Therefore, the inherent safety

- 42

(

margin and conservatisms in these methods remain unchanged." In FPE RAI 4 (Reference 41),

the NRC staff asked the licensee to clarify the safety margin discussion provided in support of

. the Approval Request. In its response (Reference 12), the licensee stated the nuclear safety analysis does not credit the Control Building Filter Unit deluge systems and the diesel fire pump and day tank room suppression systems. In addition, backup suppression is available via alternative sources. Based on the licensee's response to the RAI that the fixed systems are not credited in the fire analysis and the capability to provide alternative backup suppression from yard hydrants, the !\\IRC staff concludes that the clarification of the safety margin statements is acceptable because the safety margin credited to meet NFPA 805, Section 3.5.11, remains I

unaffected. I I

Finally, the I,icensee stated that fire protection DID will be maintained, because the configuration of the fixed systems and manual backup hose stations does not affect echelons 1, 2 and 3. The yard fire malin isolation configuration is not a fire prevention feature under echelon 1 or necessary tb ensure safe shutdown under echelon 3. Echelon 2 is maintained by the availability of automati6 detection and suppression (sprinkler system) in the Control Room HVAC Room and the availability of alternative fire brigade water sources for manual firefighting activities for the Control Room HVAC Room and the diesel fire pump and day tank rooms in the Pump house. The i water supply configuration does not compromise automatic fire detection functions or post-fire ~afe shutdown capability. Alternative hose station and hydrant connections are I

available aSlthe primary means of suppression in the event of the temporary loss of the primary water supply.

I I

Based on its review of the information submitted by the licensee, and in accordance with I

10 CFR 50.18(c)(2)(vii), the NRC staff concludes that the proposed PB method is an acceptable alternative to the corresponding NFPA 805, Section 3.5.11 requirement because it satisfies the performanc~ goals, objectives, and criteria specified in NFPA 805 related to nuclear safety and radiological1release, maintains sufficient safety margin, and maintains adequate fire protection DID.

I 3.2 Nuclear Safety Capability Assessment (NSCA) Methods I

I\\IFPA 805 (~eference 7.) is a RI/PB standard that ~lIows.engineering analyses to be used to show that F!pP features and systems provide sufficient capability to meet the requirements of 10 CFR 50.48( c) (Reference 1).

I NFPA 805, Section 2.4, "Engineering Analyses," states the following:

Engi~eering analysis is an acceptable means of eval~ating a FPP against perfdrmance criteria. Engineering analyses shall be permitted to be qualitative or quan1titative... The effectiveness of the fire protection features shall be evaluated in relbtion to their ability to detect, control, suppress, and extinguish a fire and provi~e passive protection to achieve the performance criteria and not exceed the d1amage threshold defined in Section [2.5] for the plant area being analyzed.

Chapter 1 o~ the standard defines the goals, objectives, and performance criteria that the FPP must meet il order to be in accordance with NFPA 805.

1 I

.1 I

I 1

- 43 I

j I

NFPA 805, $ection 1.3.1 "Nuclear Safety Goal":

I T~e :nuclear safety goal is to provide reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition.

I I

NFPA 805, Section 1.4.1 "Nuclear Safety Objectives":

In thk event of a fire; during any operational mode and plant configuration, the plant shall be as follows:

(11) Reactivity Control. Capable of rapidly achieving and maintaining

. subcritical conditions.

I (2) Fuel Cooling. Capable of achieving and maintaining decay heat removal I

and inventory control functions i

(p) Fission Product Boundary. Capable of preventing fuel clad damage so I

that the primary containment boundary is not challenged.

\\

I NFPA 805, Section 1.5.1 "Nuclear Safety Performance Criteria":

Fire brotection features shall be capable of providing reasonable assurance that, in the event of a fire, the plant is not placed in an unrecoverable condition. To dembnstrate this, the following performance criteria shall be met.

(i) Reactivity Control. Reactivity control shall be capable of inserting negative reactivity to achieve and maintain subcritical conditions.

Negative reactivity inserting shall occur rapidly enough such that fuel design limits are not exceeded.

(b) Inventory and Pressure Control. With fuel in the reactor vessel, head on and tensioned, inventory and pressure control shall be capable of controlling coolant level such that subcooling is maintained for a

[pressurized water reactor] (PWR) and shall be capable of maintaining or rapidly restoring reactor water level above top of active fuel for a [boiling water reactor] (BWR) such that fuel clad damage as a result of a fire is prevented.

(c) Decay Heat Removal. Decay heat removal shall be capable of removing sufficient heat from the reactor core or spent fuel such that fuel is maintained in a safe and stable condition.

(d) Vital Auxiliaries. Vital auxiliaries shall be capable of providing the necessary auxiliary support equipment and systems to assure that the systems required under (a), (b), (c), and (e) are capable of performing their required nuclear safety function.

- 44 (e) Process Monitoring. Process monitoring shall be capable oJ providing the necessary indication to assure the criteria addressed in (a) through (d) have been achieved and are being maintained.

3.2.1 Compliance with NFPA 80S NSCA Methods NFPA 80S, Section 2.4.2, "Nuclear Safety Capability Assessment," states the following:

The purpose of this section is to define the methodology for performing a NSCA.

The following steps shall be performed:

(1) Selection of systems and equipment and their interrelationships necessary to achieve the nuclear safety performance criteria in Chapter 1 (2) Selection of cables necessary to achieve the nuclear safety performance criteria in Chapter 1 (3) Identification of the location of nuclearsafety equipment and cables (4) Assessment of the ability to achieve the nuclear safety performance criteria given a fire in each fire area This section of the SE evaluates the first three steps listed above. SE Section 3.S addresses the assessment of the fourth step.

RG 1.20S, Revision 1 (Reference 8) endorses NEI 04-02, Revision 2 (Reference 9), and Chapter 3 of NEI 00-01, Revision 2, (Reference SS), and promulgates the method outlined in NEI 04-02 for conducting a NSCA. This NRC-endorsed guidance (i.e., NEI 04-02 Table B-2, "NFPA 80S Chapter 2 - Nuclear Safety Transition - Methodology Review" and NEI 00-01,.

Chapter 3) has been determined to address the related requirements of NFPA 80S, Section 2.4.2. The NRC staff reviewed LAR Section 4.2.1, "Nuclear Safety Capability Assessment Methodology," and Attachment B, "NEI 04-02 Table B Nuclear Safety Capability Assessment - Methodology Review," against these guidelines.

The endorsed guidance provided in !\\lEI 00-01, Revision 2 provides a framework to evaluate the impact of fires on the ability to maintain post-fire safe shutdown. It provides detailed guidance for:

Selecting systems and components required to meet the NSPC Selecting the cables necessary to achieve the NSPC Identifying the location of nuclear safety equipment and cables Appropriately conservative assumptions to be used in the performance of the NSCA The licensee developed the LAR based on the three guidance documents cited above.

Although RG 1.20S, Revision 1, endorses NEI 00-01, Revision 2, the licensee's review was performed to the guidance in I\\IEI 00-01, Revision 1 (Reference S6) as discussed below in SSA

- 45 RAI 2 (Reference 41). Based on the information provided in the licensee's submittal, as supplemented, a systematic process to evaluate the post-fire SSA against the requirements of NFPA 805, Section 2.4.2; Subsections (1), (2), and (3), was used, which meets the methodology outlined in the latest NRC-endorsed industry guidance.

FAQ 07-0039, "Lessons Learned - NEI 04-02 B-2 Table,'~ (Reference 33) provides one acceptable method for documenting the comparison of the SSA against the NFPA 805 requirements. This method first maps the existing SSA to the NEI 00-01, Chapter 3 methodology, which in turn, is mapped to the NFPA 805 Section 2.4.2 requirements.

The licensee performed this evaluation by comparing its SSA against the NFPA 805 NSCA requirements using the NRC endorsed process in Chapter 3 of NEI 00-01, Revision 1, and documenting the results of the review in the LAR Attachment B, "NEI 04-02 Table B-2, NFPA 805 Chapter 2 - Nuclear Safety Transition - Methodology Review," in accordance with

. the guidance of NEI 04-02.

The categories used to describe alignment with the NEI 00-01, Chapter 3 attributes are as follows:

1. The SSA directly aligns with the attribute: noted in LAR Table B-2 as "Aligns."

2. The SSA aligns with the intent of the attribute: noted in LAR Table B-2 as "Aligns with Intent."

. 3. The SSA does not align with the attribute, but there is a prior NRC approval of an alternative to the attribute, and the bases for the NRC approval remain valid: noted.in LAR Table B-2 as "Not in Alignment, but Prior NRC Approval."

4. The SSA does not align with the attribute, but there are no adverse consequences because of the non-alignment: noted in Table B-2 as "Not in Alignment, but No Adverse Consequences."

5. The SSA does not align with the attribute: noted in the B-2 Table as "Not in Alignment."

As stated above, the licensee performed the review of the NSCA to the guidance of NEI 00-01, Revision 1 instead of Revision 2. In SSA RAI 2, (Reference 41), the NRC staff requested that the licensee conduct a gap analysis to demonstrate the methodology meets the guidelines of NEI 00-01, Revision 2. In its response to SSA RAI 2 (Reference 13), the licensee stated that a gap analysis was performed and it indicated that there are no significant differences between the revisions. The licensee further stated that based on its review against the endorsed criteria, the plant aligns with the guidance provided in NEI 00~01, Revision 2.

As described in the LAR, the licensee's process for identification and eyaluation of MSOs used an expert panel and followed the guidance of NEI 04-02, Regulatory Guide 1.205 and FAQ 07-0038, "Lessons Learned on Multiple Spurious Operations" (Reference 32).

- 46 The expert panel used by the licensee consisted of subject matter experts with experience in electrical engineering, FPRA, PRA, safe shutdown analysis, fire protection, and plant operations.

With respect to instrumentation tubing credited for safe shutdown (NEI 00-01, Section 3.2.1.1, 3.2.1.5 and 3.2.1.7), a review was conducted and additional discussion is provided in Section 3.2.1.1 below.

3.2.1.1 Attribute Alignment - Aligns For the majority of the NEI 00-01, Chapter 3 attributes, *the licensee determined that the SSA aligns directly with the attribute. In these instances, based on the validity of the licensee's statements, the NRC staff concludes that the licensee's statements of alignment are acceptable.

The following attributes were identified in the LAR Attachment Bf Table B-2, as aligning via this method and requiring additional review by the NRC staff:

3.2.1.2 Fire Damage to Mechanical Components (not electrically supervised) 3.2.1.5 Instrument Failures 3.2.1.7 Instrument Tubing LAR Attachment B, Table B-2 states that the licensee's methodology aligns with Sections 3.2.1.2,3.2.1.5, and 3.2.1.7, of NEI 00-01, Revision 1. In NEI 00-01, Revision 2, each section provides additional guidance, regarding instrumentation tubing credited for safe shutdown. In SSA RAI 3 (Reference 41), the !\\IRC staff requested additional information from the licensee regarding instrumentation tubing failure modes. In its response to SSA RAI 3, (Reference 12), the licensee stated that heat sensitive piping is assumed to fail resulting in loss of associated instrument function. The licensee further stated that heat sensitive piping is identified as that constructed with brazed or soldered joints per the guidance of NEI 00-01, Revision 2, Section 3.2.1.2. The licensee further stated that design documents including Field Sketch Drawings provide installation details and specify materials of construction for instrument tubing and that the pertinent bills of material for instrument tubing within the scope of the NSCA show the tubing material as ferrous with joints that are not brazed or soldered.

Therefore, the NRC staff determin~d that, based on the information provided in LAR Table B-2, the LAR supplements, and the NRC staff review, the licensee meets the requirements.

3.2.1.2 Attribute Alignment - Aligns with Intent In several of the NEI 00-01, Chapter 3 attributes the licensee determined that the SSA aligns with the intent of the attribute, and provided additional clarification when describing its means of alignment. The attributes identified in LAR Attachment B, Table B-2, as having this condition are as follows:

3; Deterministic methodology.

3.1.A; Safe Shutdown Systems and Path Development 3.1.1.8; Safety-Related Equipment 3.1.1.9; 72 Hour Coping 3.1.2.4; Decay Heat Removal

- 47 I\\)

3.1.3.3; Define Combinations of Systems for Each Safe Shutdown Path e

3.1.3.4; Assign Shutdown Paths to Each Combination of Systems

• 3.2.1.4; Check Valves

• 3.2.2.1; Identify the System Flow Path for Each Shutdown Path

• 3.2.2.2; Identify the Equipment in Each Safe Shutdown System Flow Path Including Equipment That May Spuriously Operate and Affect System Operation

• 3.2.2.3; Develop a List of Safe Shutdown Equipment and Assign the Corresponding System and Safe Shutdown Path(s) Designation to Each.

• 3.2.2.4; Identify Equipment Information Required for the Safe Shutdown Analysis

• 3.3.1.1; General Fire Prevention Activities

• 3.3.1.2; Control of Combustible Materials

• 3.4.1.5; Repairs

• 3.5.1.3; Duration of Circuit Failures Achieving and Maintaining Fuel in a Safe and Stable Condition (3.1.1.9. 3.1.2.4. 3.2.2.4)

Three of the attributes of NEI 00-01 (3.1.1.9, 3.1.2.4, and 3.2.2.4) for which the licensee stated alignment with intent, address the capability for achieving and maintaining fuel in a safe and stable condition and the licensing basis is to achieve and maintain hot shutdown (Mode 3) conditions. In SSA RAI 4 (Reference 41), the NRC staff requested additional information regarding long term actions required to maintain safe and stable conditions. In response to SSA RAI 4 (Reference 12). the licensee stated that in the hot shutdown state, decay heat removal is accomplished by allowing steam to flow from the reactor pressure vessel (RPV) to either the main condenser or the suppression pool. If the main condenser is used, heat is rejected to the atmosphere via large cooling towers, but if the suppression pool is used, heat is rejected to the Cedar River via one or both of the RHR heat exchangers. The licensee stated that the first method relies partially on non-safety related equipment, such as circulating water, condensate, and feedwater pumps, while the second method relies solely on safety-related equipment. The licensee further stated that in the event that offsite power sources are not available, the second

. method is used since the safety related equipment is powered by on-site emergency diesel generators. The licensee further stated that the NSPC RPV inventory function can be met with either high pressure or low pressure injection systems and that viable high pressure systems are Feedwater, High Pressure Coolant Injection (HPCI), Reactor Core Isolation Cooling (RCIC),

and Control Rod Drive (CRD), while viable low pressure systems are Condensate, Low Pressure Coolant Injection (LPCI), and Core Spray.

The licensee stated that the Cedar River serves as a reliable, constant source of water for supporting decay heat removal and component cooling needs and that if the main condenser is used as the primary heat rejection mechanism, river water is pumped from the intake structure to the Circ Water Pit at a rate sufficient to make up for cooling tower water vapor losses. The licensee further stated that if suppression pool cooling is used as the primary heat rejection mechanism, river water is supplied at a rate equivalent to the flow rate of operating RHR Service Water and Emergency Service Water pumps.

The licensee stated that the Condensate Storage Tanks (CSTs) provide a source of water for operation of the HPCI and RCIC systems and that if the tanks become depleted, the suction source for both HPCI and RCIC can be transferred to the suppression pool. The licensee further stated that in this mode, water loss from the primary containment is minimized by

- 48 directing steam from the RPV to the suppression pool and that the CSTs also provide makeup to the hotwell in the event that the main condenser is being used for decay heat removal. The licensee further stated that only a small amount of water is needed for this purpose however, since only water lost due to normal system leakage needs to be replaced.

The licensee stated that pressure control of the RPV is accomplished by operation of Safety Relief Valves (SRVs) and that the SRVs rely on pressurized nitrogen for their operation, which is supplied by the Orywell Pneumatics system. The licensee further stated that four large accumulators provide pneumatic pressure for cycling SRVs and that nitrogen for makeup is available on-site and is provided to the Drywell Pneumatics system automatically by the Nitrogen Makeup System. The licensee further stated that the Nitrogen Makeup System is assumed in the PRA program to contain sufficient nitrogen for successful operation of drywell pneumatic loads for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and that in the longer term, additional nitrogen may need to be delivered to the site. The licensee further stated that ordering and taking delivery of nitrogen is a routine activity and ample time would be available for achieving this action in the event that supplies rU,n low while operating in hot shutdown mode.

The licensee stated that if one or both of the standby diesel generators (SBOGs) are operated for on-site AC power supply, delivery of additional fuel oil would be needed if the normal seven day supply runs low and that ordering and taking delivery of diesel fuel oil is a routine activity and w!th a seven day supply already available on site, ample time would be available for achieving this action if needed.

The licensee stated that adequate negative reactivity for the prevention of re-criticality while in the hot shutdown mode is provided by inserted control rods and that in the unlikely event that one or more control rods fail to insert, the standby liquid control (SBLC) system can be manually operated to maintain the fuel in a sub-critical condition. The licensee further stated that

additional supplies of boron are' kept on site in the unlikely event SBLC fails to inject when called upon and that the risk impact of needing to obtain commodities such as nitrogen, diesel fuel oil, and boron is very low based on the long period of time before the depletion of such items becomes a concern, and on the routine nature of ordering and taking delivery of them.

The licensee further stated that for fire initiated events where the Emergency Response Organization is activated, additional personnel are available for initiating these tasks, further ensuring they will be reliably accomplished. The NRC staff concludes that the methods as described by the licensee are acceptable because they are sufficiently similar to the specific methods in NEI 00-01, and therefore align with the intent of NRC endorsed guidance.

Remaining NEI 00-01 Attributes The remaining NEI 00-01 attributes for which the licensee stated aligns with intent (3, 3.1.A, 3.1.1.8, 3.1.3.3, 3.1.3.4, 3.2.1.4, 3.2.2.1, 3.2.2.2, 3.2.2.3, 3.3.1.1, 3.3.1.2, 3.5.1.3, and 3.4.1.5),

describe similar means or methods that were applied to achieve the intended result of the NEI 00-01 guidance. The NRC staff concludes that the methods as described by the licensee are acceptable because they are sufficiently similar to the specific methods in NEI 00-01, and therefore align with the intent of NRC endorsed guidance.

- 49 3.2.1.3 Attribute Alignment - !\\lot in Alignment, but Prior NRC Approval The licensee did not identify any attributes in this category.

3.2.1.4 Attribute Alignment - Not in Alignment, but No Adverse Consequences The licensee did not identify any attributes in this category.

3.2.1.5 Attribute Alignment - !\\lot in Alignment The licensee did not identify any attributes in this category.

3.2.1.6 NFPA 805 NSCA Methods Conclusion The NRC staff reviewed the documentation provided by the licensee describing the process used to perform the NSCA required by NFPA 805, Section 2.4.2. The licensee performed this evaluation by comparing the SSA against the NFPA 805 NSCA requirements using NEI 00-01, Revision 1 with a gap analysis to the NRC-endorsed process in Chapter 3 of NEI 00-01, Revision 2. The results of the review are documented in LAR Attachment B, Table B-2, in accordance with NEI 04-02, Revision 2 and the gap analysis of NEI 00-01, Revision 2, is discussed in SSA RAI 2 (Reference 41). The licensee indicated that there are no significant differences between alignment with NEI 00-01, Revision 1 and NEI 00-01, Revision 2.

Based on the information provided in the licensee's submittal, as supplemented, the NRC staff accepts the method the licensee used to perform the NSCA with respect to the selection of systems and equipment, selec.tionof cables, and identification of the location of nuclear safety equipment and cables, as required by NFPA 805, Section 2.4.2. The NRC staff concluded that the licensee's method is acceptable because it either:

Met the NRC-endorsed guidance directly.

Met the intent of the endorsed guidance with adequate justification.

3.2.2 Maintaining Fuel in a Safe and Stable Condition The nuclear safety goals, objectives and performance criteria of NFPA 805 allow more flexibility than the previous deterministic requirements based on Appendix R to 10 CFR 50 (Reference 4) and NUREG-0800, Section 9.5.1 (Reference 57), since NFPA 805 only requires the licensee to maintain the fuel in a safe and stable condition, rather than achieve and maintain cold shutdown in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In LAR Section 4.2.1.2, the licensee stated that the NFPA 805 licensing basis is to achieve and maintain the fuel in hot shutdown (Mode 3) conditions following any fire occurring with the reactor operating at power, shutdown prior to aligning the RHR system for shutdown cooling, or in transition between these two operational phases.

For the most limiting fire scenarios in every fire area, the licensee has documented the availability of long term decay heat removal provided by water from the torus, with temperature maintained by RHR operating in the suppression pool cooling mode. The significant volume of water in the torus is available for primary makeup to match nominal system losses.

- 50 The licensee stated that initiation of RHR in the suppression pool cooling mode does not imply that the plant would proceed all the way to cold shutdown and that following stabilization at hot shutdown, a long term strategy for reactivity control, decay heat removal, and inventoryl pressure control would be determined based on the extent of equipment damage. The licensee

. further stated that if an assessment of the post-fire conditions indicated that placing RHR in the shutdown cooling mode would be advisable, then repair activities would commence in a safe and controlled manner to restore plant equipment necessary for reactor cooldown.

On the basis of the licensee's analysis as described in the LAR, as supplemented, the NRC staff concludes that the licensee has provided reasonable assurance that the fuel can be maintained in a safe and stable condition, post-fire, for an extended period of time.

3.2.3 Applicability of Feed and Bleed The limitations of 10 CFR 50.48(c)(2)(iii), "Use offeed-and-bleed," are not applicable to BWRs.

3.2.4 Assessment of Multiple Spurious Operations NFPA 805, Section 2.4.2.2.1, "Circuits Required in Nuclear Safety Functions" states that:

Circuits required for the nuclear safety functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. [Nuclear Safety Capability Systems and Equipment Selection] This evaluation shall consider fire induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and Signals.

In addition, NFPA 805, Section 2.4.3.2, states that the PSA evaluation shall address the risk contribution associated with all potentially risk-significant fire scenarios. Because the RI/PB approach taken used FREs in accordance with NFPA 805 Section 4.2.4.2, "Use of Fire Risk Evaluation," adequately identifying and including potential MSO combinations is required to ensure that all potentially risk-significant fire scenarios have been evaluated.

The NRC staff reviewed LAR Section 4.2.1.4, "Evaluation of Multiple Spurious Operations," and LAR Attachment F, "Fire-Induced Multiple Spurious Operations Resolution," to determine whether the licensee has adequately addressed MSO concerns. As described in the LAR, the licensee's process for identification and evaluation of MSOs used an expert panel and followed the guidance of NEI 04-02, RG 1.205, and FAQ 07-0038, (Reference 32). The expert panel consisted of subject matter experts with experience in electrical engineering, fire PRA, PRA, SSA, FPE, and plant operations.

LAR Attachment F stated that the licensee conducted an initial expert panel review in 2008 and a second review in 2010. Prior to initial review, the panel was provided with training and was provided with a specific project instruction for conducting the review. The expert panel sources for information and identifying MSOs included the SSA, generic lists (e.g. from Owners Groups),

- 51 self-assessment results, and internal events PRA insights. The NSCA and Fire PRA were updated to reflect the treatment of applicable MSO scenarios. This included the identification of equipment, cables, and cable routing by plant locations. The MSO combination components of concern were also evaluated as part of the NSCA. For cases where the pre-transition MSO combination components did not meet the deterministic compliance, the MSO combination components were added to the scope of the FREs.

LAR Attachment F, "Fire-Induced Multiple Spurious Operations Resolution," describes the process the licensee utilized to address MSOs. That process includes 5 steps: 1. Identify potential MSOs of concern; 2. Conduct an expert panel to assess plant specific vulnerabilities;

3. Update the Fire PRA model and NSCA to include the MSOs of concern; 4. Evaluate for NFPA 805 Compliance; and, 5. Document Results. As described in LAR Attachment F, under the results for Steps 3, 4, and -5, the MSOs identified in Steps 1 and 2 were incorporated in the fire PRA model and evaluated for inclusion in the NSCA. Variances from Deterministic Requirements (VFDRs)'were created where MSO combinations did not meet the deterministic requirements of NFPA 805, Section 4.2.3 and these VFDRs were addressed using the PB approach of NFPA 805, Section 4.2.4. Based on the evaluations, components associated with the MSOs were added to the NSCA equipment list and logics, and cable tracing and circuit analysis was performed. The fire PRA quantified the fire-induced risk model containing the MSO pathways. The MSO contribution is included in the fire PRA results, including those associated with VFDRs in the FREs.

The NRC staff reviewed the expert panel process for identifying circuits susceptible to MSOs as described above and concludes that the licensee adopted a systematic and comprehensive process for identifying MSOs to be analyzed utilizing available industry guidance. Furthermore, the process used provides reasonable assurance that the FRE appropriately identifies and includes risk significant MSO combinations. Based on these conclusions, the NRC staff concludes that the licensee's approach for assessing the potential for MSO combinations is acceptable.

3.2.5 Establishing Recovery Actions I\\IFPA 805, Section 1.6.52, "Recovery Action," defines an RA as follows:

Activities to achieve the nuclear safety performance criteria that take place outside the main control room or outside the primary control station(s) for the equipment being operated, including the replacement or modification of components.

NFPA 805, Section 4.2.3.1 states that:

One success path of required cables and equipment to achieve and maintain the nuclear safety performance criteria without the use of recovery actions shall be protected by the requirements specified in either 4.2.3.2, 4.2.3.3, or 4.2.3.4, as applicable. Use of recovery actions to demonstrate availability of a success path for the nuclear safety performance criteria automatically shall imply use of the performance-based approach as outlined in 4.2.4.

- 52 NFPA 805, Section 4.2.4, "Performance-Based Approach," states the following:

When the use of recovery actions has resulted in the use of this approach, the additional risk presented by their use shall be evaluated.

The NRC staff reviewed LAR Section 4.2.1.3, "Establishing Recovery Actions," and Attachment G, "Recovery Actions Transition," to evaluate whether the licensee meets the associated requirements for the use of RAs per NFPA 805.

The licensee used the endorsed guidance provided in NEI 04-02, Section 4.6 and the guidance in FAQ 07-0030 (Reference 31) to establish the population of RAs being carried forward in the RI/PB FPP. RAs addressed during the NFPA 805 transition process included the consideration of existing operator manual actions (OMAs) in the deterministic FPP, as well as those being added based on the VFDRs identified in the individual fire area assessments. OMAs are actions performed by plant operators to manipulate components and equipment from outside the main control room to achieve and maintain postfire hot shutdown, not including "repairs." Operator manual actions include an integrated set of actions needed to ensure that hot shutdown can be.

accomplished for a fire in a specific plant area. OMAs are transitioned to Recovery Actions (RAs) under NFPA 805. Recovery actions are activities to achieve the nuclear safety performance criteria that take place outside of the main control room or outside of the primary control station(s) for the equipment being operated, including the replacement or modification of components.

There is one location designated as an alternate shutdown panel (ASP) and four locations designated as primary control stations (PCSs) as defined in RG 1.205. In SE dated January 6, 1983, (Reference 58) the NRC staff documented its evaluation of the alternate shutdown capability design and concluded it to be in accordance with Appendix R to 10 CFR 50, Items III.G.3 and IILL (Reference 4).

DAEC utilized the guidance in RG 1.205, Revision 1 (Reference 8) for addressing recovery actions. This included consideration of the definition of PCS and recovery action, as clarified in the RG 1.205, Revision 1. Accordingly, any actions required to transfer control to, or operate equipment-from, the PCS, while required as part of the RI/PB FPP, were not considered recovery actions per the RG 1.205 guidance and in accordance with NFPA 805. Alternatively, any operator manual actions required to be performed outside the control room and not at the PCS were considered recovery actions.

In SSA RAI 9. (Reference 41), the NRC staff requested that the licensee identify the dampers and describe the process of blocking open fire dampers for the two RAs for the 1A Switchgear Room Air Supply and Exhaust Fans. In its response to SSA RAI 9 (Reference 12), the licensee stated that the referenced RAs as part of the steps required for establishing alternate switchgear room ventilation and that the inclusion of damper operation was a typographical error in the VFDR which carried through into Attachment G of the LAR. The licensee further stated that no dampers require operation to establish alternate switchgear room ventilation.

OMAs meeting the definition of a RA are required to comply with the NFPA 805 requirements outlined above. Some OMAs may not be required to demonstrate the "availability of a success path," in accordance with NFPA 805, Section 4.2.3.1, but may still be required to be retained in

~ 53 the RI/PB FPP because of DID considerations as described in Section 1.2 of NFPA 805. The licensee did not differentiate between an RA that is needed to meet the NSCA and one retained to provide DID. In each instance, the licensee determined whether a transitioning OMA was an RA or not necessary for the post-transition RI/PB FPP.

The licensee stated that all credited RAs, as listed in LAR Attachment G, were subjected to a feasibility review. In accordance with the NRC-endorsed guidance in NEI 04-02, the feasibility criteria used in the licensee's assessment process, were based on the criteria described in FAQ 07-0030 (Reference 31) and each of the 11 individual feasibility attributes were addressed.

LAR Attachment G, Table G-1, "Recovery Actions and Activities Occurring at the Primary Control Stations," describes each RA associated with disposition of a VFDR from the fire area assessments as documented in LAR Attachment C, "Fire Area Transition." The feasibility review was based on documentation only, including previous feasibility evaluations for safe shutdown OMAs. The licensee included Implementation Item 12 in Table S-2 to revise post-fire safe shutdown procedures and training as necessary to incorporate updated NSCA strategies.

Based on the above. considerations, the NRC staff concludes that the licensee has followed the endorsed guidance of NEI 04-02 and RG 1.205 to identify and evaluate RAs in accordance with NFPA 805, and therefore, there is reasonable assurance of meeting the regulatory requirements of 10 CFR 50.48(c). The NRC staff concludes that the feasibility criteria applied to RAs are acceptable based on conformance with the endorsed guidance contained in NEI 04-02 and successful completion of identified Implementation Item 12 in Table S-2.

3.2.6 Conclusion for Section 3.2.

The NRC staff reviewed the licensee's LAR, as supplemented, for conformity with the requirements contained in NFPA 805, Section 2.4.2, regarding the process used to perform the NSCA. The NRC staff concluded that the declared safe and stable condition proposed was acceptable and that the licensee's process is adequate to appropriately identify and locate the systems, equipment, and cables, required to provide reasonable assurance of achieving and maintaining the fuel in a safe and stable condition,. as well as to meet the NFPA 805 NSPC.

The NRC staff also reviewed the licensee's process to identify and analyze MSOs. Based on the LAR, as supplemented, the process used to identify and analyze MSOs is considered comprehensive and thorough. Through the use of an expert panel process, in accordance with the guidance of RG 1.205, NEI 04-02, and FAQ 07-0038, potential MSO combinations were identified and included as necessary in the NSCA, as well as the applicable FREs. The NRC staff also considers the approach the licensee uses for assessing the potential for MSO combinations acceptable, because it was performed in accordance with NRC-endorsed guidance.

Pending completion of Implementation Item 12 in Attachment S, Table S-2 of the LAR, the NRC staff concludes that the process used by the licensee to review, categorize, and address RAs during the transition is consistent with the NRC-endorsed guidance contained in NEI 04-02 and RG 1.205. Therefore, the information provided by the licensee provides reasonable assurance that the regulatory requirements of 10 CFR 50.48(c) and NFPA 805 for NSCA methods are met.

- 54 3.3 Fire Modeling NFPA 805 (Reference 7) allows both, fire modeling and FREs as PB alternatives to the deterministic approach outlined in the standard. These two PB approaches are described in NFPA 805, Sections 4.2.4.1 and 4.2.4.2, respectively. Although fire modeling and FRE are presented as two different approaches for PB compliance, the FRE approach generally involves some degree of fire modeling to support engineerin'g analyses and fire scenario development.

NFPA 805, Section 1.6.18, defines a fire model as a "mathematical prediction of fire growth,

, environmental conditions, and potential effects on structures, systems, or components based on the conservation equations or empirical data."

The NRC staff reviewed LAR (Reference 10) Section 4.5.2, "Performance-Based Approaches,"

which describes how the licensee used fire modeling as part of the transition to NFPA 805 at DAEC, and LAR Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805," which describes how the licensee performed fire modeling calculations in compliance with the NFPA 805 PB evaluation quality requirements for fire protection systems and features at DAEC, to determine whether the fire modeling used to support transition to NFPA 805 is acceptable.

In LAR Section 4.5.2.1, the licensee indicated that the fire modeling approach (NFPA 805 Section 4.2.4.1) was not used for the DAEC NFPA 805 transition. The licensee used the FRE PB method (i.e., fire PRA) with input from fire modeling analyses. Therefore, the NRC staff reviewed the technical adequacy of the DAEC FRE, including the supporting fire modeling analyses, as documented in Section 3.4.2 of this SE~ to evaluate compliance with the NSPC.

The licensee did not propose any fire modeling methods to support PB evaluations in accordance with NFPA 80S, Section 4.2.4.1 t as the sole means for demonstrating compliance with the NSPC. There are no plant-specific fire modeling methods acceptable for use to support compliance with NFPA 805, Section 4.2.4.1, as part of this licensing action supporting the transition to NFPA 805 at DAEC.

3.4' Fire Risk Assessments This section addresses the licensee's fire risk evaluation performance-based method, which is based on NFPA 805, Section 4.2.4.2. The licensee chose to use only the fire risk evaluation performance-based method in accordance with NFPA 80S, Section 4.2.4.2. Th'e fire modeling performance-based method of NFPA805 Section 4.2.4.1 was not used for this application.

NFPA 80S, Section 4.2.4.2, "Use of Fire Risk Evaluation," states the following:

Use of fire risk evaluation for the performance-based approach shall consist of an integrated assessment of the acceptability of risk, defense in depth [DID], and safety margins.

The evaluation process shall compare the risk associated with implementation of the deterministic requirements with the proposed alternative. The difference in

. risk between the two approaches shall meet the risk acceptance criteria described in [NFPA 805,] Section 2.4.4.1 ["Risk Acceptance Criteria"]. The fire

- 55 risk shall be calculated using the approach described in NFPA 805, 2.4.3 ["Fire Risk Evaluations"].

3.4.1 Maintaining Defense-in-Depth and Safety Margins NFPA 805, Section 4.2.4.2, requires that the "use of fire risk evaluation for the performance based approach shall consist of an integrated assessment of the acceptability of risk, defense in-depth, and safety margins."

3.4.1.1 Defense-in-Depth (DID)

NFPA 805, Section 1.2, states the following:

Protecting the safety of the public, the environment, and plant personnel from a plant fire and its potential effect on safe reactor o.perations is paramount to this standard. The fire protection standard shall be based on the concept of defense in-depth. Defense-in-depth shall be achieved when an adequate balance of each of the following elements is provided:

• Preventing fires from starting

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

• Providing an adequate level of fire protection for structures, systems, and components important to safety, so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed The NRC staff reviewed LAR Section 4.8.1, "Results of the Fire Area Review," and LAR Attachment C Table B-3, "NEI 04-02 Table B Fire Area Transition," as well as the associated supplemental information, in order to determine whether the principles of DID were maintained in regard to the planned transition to NFPA 805 at DAEC.

The licensee developed a methodology for evaluating DID which defines each of the thre.e DID elements identified in the LARAttachment L, and in more detail in the response to PRA RAI 63 (Reference 16), referred to as echelons 1, 2, and 3, respectively. The licensee provided a table where, for each of the three echelons, several examples of fire protection features that addressed that echelon are identified, along with a discussion of the considerations used in assessing those features. The assessment determined whether changes would be needed to assure that each\\ echelon has been satisfactory achieved or whether reliance on features in other echelons were needed and should be developed. Many of the identified fire protection features are required to be in place in order to demonstrate compliance with the fundamental fire protection program and design elements of NFPA 805 Chapter 3 (e.g., combustible control program, hot work control program, etc.). However, the capabilities for some of the fire protection features for DID were evaluated and improved as needed based on the results of the performance-based analyses conducted during the NFPA 805 transition (e.g., detection system, suppression system, ERFBS, use of fire rated cable, use of RAs, etc.).

- 56 As described in the response to PRA RAI 63, this method for addressing defense-in-depth was implemented in the fire risk evaluations (FREs) performed on each performance-based fire area. The licensee stated that the FREs evaluate VFDRs using an integrated assessment of risk, DID, and s*afety margins. DAEC evaluated the VFDRs and fire area risk and scenario consequences to identify general DID echelon imbalances. The licensee further stated that potential methods to balance the DID features were identified to ensure an adequate balance of DID features is maintained for the Fire Area. Finally, the licensee stated that the results of the FRE reviews associated with DID identified the need to implement and credit recovery actions, as well as suppression and detection systems.

Based on its review of the response to PRA RAI 63, and the FREs during its audit of the DAEC NFPA 805 transition to RI/PB FPP, the NRC staff concludes that the licensee has systematically and comprehensively evaluated fire hazards, area configuration, detection and suppression features, and administrative controls in each fire area and concludes that the methodology as proposed in its LAR adequately evaluates DID against fires as required by NFPA 805 and therefore the proposed RI/PB FPP adequately maintains DID.

3.4.1.2 Safety Margins NFPA 805, Section 2.4.4.3 states the following:

The plant change evaluation shall ensure that sufficient safety margins are maintained.

NEI 04-02, Section 5.3.5.3, "Safety Margins," lists two specific criteria that should be addressed when considering the impact of plant changes on safety margins:

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

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

LAR Section 4.5.2.2, "Fire Risk Approach," states that safety margins were considered as part of the FRE process and that this process is based on the req)Jirements of NFPA 805, industry guidance in NEI 04-02, and RG 1.205. An FRE was performed for each fire area containing VFDRs.The FREs contain the details of the licensee's review of safety margins for each performance-based fire area.

In its response to PRA RAI 62 (Reference 16), the licensee further described the methodology used to evaluate safety margins in the FREs to include the following evaluations and determinations:

• The development of the PRA logic model and the fire PRA application were performed in accordance with industry accepted codes and standards including 10 CFR 50.48(c),

NFPA 805 (2001 edition), ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S 2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk

- 57 Assessment for Nuclear Power Plant Applications," (Reference 59) and RG 1.200, Revision 2 (Reference 26).

• The Fire PRA, including fire modeling performed in support of the Fire PRA, was developed utilizing industry, NRC, and National Institute of Standards and Technology accepted codes and standards including NUREG/CR-6850, NEI 04-02, associated Frequently Asked Questions resolutions, and Fire PRA methods used to support the LAR.

• Plant system performance parameters (e.g., heat transfer coefficients, pump performance curves, etc.), originally established to meet the nuclear safety performance criteria contained in the DAEC accident analyses, were not modified in the PRA or FRE.

The Fire PRA methods used to support the LAR were evaluated by the NRC staff in Section 3.4.2.2 of this SE, and the NRC staff did not accept some of the methods used to support the LAR. Fire PRA methods that are not accepted by the NRC are not an alternative to NRC accepted codes and standards. In each case, the licensee removed the method from the PRA or provided an implementation item in the LAR Attachment S, as confirmed in the response to PRA RAI 84 (Reference 21).

The safety margin criteria described in NEI 04-02, Section 5.3.5.3 and the LAR, as supplemented, are consistent with the criteria as described in RG 1.174 and therefore acceptable. DAEC used appropriate codes and standards, and has removed the unacceptable methods, or has committed to updating the Fire PRA. Based on its review of the LAR and the response to PRA RAI 62, and the FREs during its audit of the DAEC NFPA 805 transition RI/PB FPP, the NRC staff concludes that the licensee's approach has adequately addressed the issue of safety margins in the implementation of the fire risk evaluation process.

3.4.1.3 Defense-in-Depth and Safety Margin Conclusion Based on the information provided by the licensee in the LAR, as supplemented, the transition process included a detailed review of fire protection DID and safety margins. The individual*

FRE, LAR Table 4-3, and LAR Attachment C Table B-3 document the results of the DID and safety margins review. The NRC staff concludes that the licensee's evaluation in regard to DID and safety margins are acceptable, because the licensee's process and results followed the endorsed guidance in NEI 04-02, Revision 2, and are consistent with the NRC staff guidance in RG 1.205, Revision 1, and RG 1.174, Revision 1. Section 3.5 of this SE discusses the results of the individual fire area reviews, including the documentation of the required suppression and detection systems.

- 58 3.4.2 Quality of the Fire Probabilistic Risk Assessment The objective of the PRA quality review is to determine whether the plant-specific PRA used in evaluating the proposed LAR is of sufficient scope, level of detail, and technical adequacy for the application. The NRC staff evaluated the PRA quality information provided by the licensee in its NFPA 805 submittal, as supplemented, including industry peer review results and self assessments performed by the licensee. The NRC staff reviewed LARSection 4.5.1, "Fire PRA Development and Assessment," Section 4.7, "Program Documentation, Configuration Control, and Quality Assurance," Attachment C, "NEI 04-02 Table B Fire Area Transition,"

Attachment U, "Internal Events PRA Quality," Attachment V, "Fire PRA 'Quality," and Attachment W, "Fire PRA Insights."

The licensee developed its internal events PRA during the Individual Plant Examination process and continued to maintain and improve the PRA as Regulatory Guide 1.200, "An Approach For Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities," and supporting industry standards have evolved. The licensee developed its Fire PRA model for both LeVel 1 (core damage) and partial Level 2 (large early release) PRA during at-power conditions. For the development of the Fire PRA, the licensee modified its internal events PRA model to capture the effects of fire.

The licensee did not identify any: (1) known outstanding plant changes that would require a change to the Fire PRA model, or (2) any planned plant changes that would significantly impact the PRA model, beyond those identified and scheduled to be implemented as part of the transition to a FPP based on NFPA 805. Based on this information, the NRC staff concludes that the Fire PRA model for DAEC meets the criteria, that it represents the current, as built, as operated configuration, and is therefore capable of being adapted to model both the post transition and compliant plant as needed.

The licensee identified administrative controls and processes used to maintain the Fire PRA.

model current with plant changes and to evaluate any outstanding changes not yet incorporated into the PRA model for potential risk impact as a part of the routine change evaluation process.

Further, as described in Section 3.8.3 of this safety evaluation, the licensee has a program for ensuring that developers and users of these models are appropriately trained and qualified.

Th'erefore the NRC staff concludes that the PRA should be capable of supporting post-transition FREs to support, for example, the self-approval process, after any changes required during implementation are completed.

3.4.2.1 Internal Events PRA Model The licensee's evaluation of the technical adequacy of the portions of its internal events PRA model used to support development of the Fire PRA model included a combination of peer reviews and gap assessments. The DAEC internal events PRA full scope peer review was performed in December 2007 using the NEI 05-04 process, the combined PRA standard, ASME/ANS-RA-Sa-2005, and RG 1.200, Revision 1. A focused scope peer review of the internal events PRA was conducted in March 2011 using the combined standard, ASME/ANS RA-Sa-2009, and RG 1.200, Revision 2. A gap assessment was also performed, as given in the response to PRA RAI 65 (Reference 17), for the supporting requirements (SRs) not within the scope of the focused scope peer review by comparing the SRs in the two combined

- 59 standards. SRs are detailed, focused statements of "good PRA practice" which, collectively, comprise what is deemed satisfactory for a technically adequate PRA. The internal events PRA model that was reviewed for the focused scope peer review serves as the basis of the Fire PRA used in performing PRA evaluations for the LAR.

Within each SR, there are three degrees of "satisfaction" - these are the Capability Categories.

Three are common (I, II, and III), with I being the minimum, II is considered widely acceptable, and III going beyond the state of the art. For each Supporting Requirement, a PRA reviewer (in the peer review) assigns one of these Capability Categories.

Attachment U of the LAR provides the licensee's dispositions to all 12 facts and observations (F&Os). In general, an F&O is written for any SR if that SR does not fully satisfy Capability Category II of the ASME standard, consistent with RG 1.200, Revision 2.

As described in Attachment U, the licensee dispositioned each F&O by assessing the impact of the F&O on the Fire PRA and the results for the NFPA-805 application. The NRC staff requested additional information to assess the adequacy of some of the F&O dispositions for

. the review. The NRC staff evaluated each F&O and the licens.ee's disposition in Attachment U that the F&O did not have any significant impact for the application. The NRC staff's review and conclusion for DAEC's resolution of each of F&O is summarized in the NRC's Record of Review dated July 31, 2013 (Reference 60).

As a result of the review of the LAR, as supplemented, the NRC staff concludes that the DAEC internal events PRA is sufficiently technically adequate because its quantitative results, considered together with sensitivity study results, can be used to demonstrate that the change in risk due to the transition to NFPA-805 meets the acceptance guidelines of RG 1.174 and are acceptable. To reach this conclusion, the NRC staff has reviewed all F&Os provided by the peer reviewers and determined that the resolution of every F&O supports the determination that the quantitative results are adequate. Accordingly, the NRC staff concludes that the licensee has demonstrated that the internal events PRA meets the guidance in RG 1.200, Revision 2, that it is reviewed against the applicable supporting requirements (SRs) in ASME/ANS-RA-Sa 2009, and that it is technically adequate to support the fire risk evaluations and other risk calculations required for the NFPA 805 application.

3.4.2.2 Fire PRA Model The licensee evaluated the technical adequacy of the DAEC Fire PRA model by conducting a peer review of the Fire PRA model using the SRs of ASME/ANS-RA-Sa-2009 (Reference 59) as endorsed by RG 1.200, Revision 2 (Reference 26). The full scope peer review of the Fire PRA was performed in June 2010, and assigned a Capability Category assessment to each SR in the Fire PRA element. The licensee addressed F&Os related to incorporation of fire related elements into the internal events PRA. In addition, the licensee assessed F&Os from the March 2011 internal events PRA focused scope peer review for their impact on the Fire PRA as given in the LAR, Attachment U.

Table V-1 of Attachment V in the LAR provides the Fire PRA peer review and licensee assessment F&Os against SRs that were met, not met, achieved CapabilityCcategory I, II, or III, or were not applicable. Table V-2 of Attachment V in the LAR provides SRs against which*

- 60 DAEC proposed that Capability Category I, or a deferred F&O resolution due to methods under industry review, was sufficient for the application. Table V-3 in Attachment V in the LAR provides the licensee's disposition of the F&Os.

The NRC staff reviewed the licensee's dispositions to all of the F&Os to determine the technical adequacy of the fire events PRA for the NFPA 805 application. The NRC staff's review and conclusion for the licensee resolution of each of the F&Os is summarized in a record of review (Reference 61). The NRC staff requested additional information for the review of some of the F&Os, and issues identified are discussed below. In addition, a discussion is included on Fire PRA modeling aspects and risk insights relevant to the application, and sensitivity studies performed in resolving Fire PRA methods issues are discussed below.

A sensitivity study is the common practice in both PRA and other engineering analyses where the value of a parameter that is used for a quantitative evaluation is varied between its extremes (low and high) without any other parameters being varied at the same time. This enables determination of how much the result of the evaluation is affected by thE? potential variability in the parameter. Section 3.4.7 of this SE discusses additional sensitivity analyses relevant to the application.

A transient fire HRR had been used in the Fire PRA that is lower than that provided in NUREG/CR-6850 (Reference 62) guidance. Section V2.2 of the LAR and F&O 3-9 identified that a 98% 317kW transient fire was replaced with a 69kW transient fire. In PRA RAI 1.01 (Reference 64), the NRC staff requested that a sensitivity analysis be performed using the NUREG/CR-6850 transient HRR guidance unless a lower HRR could be justified. The response to PRA RAI 1.01 (Reference 18) showed that using the larger HRR in NUREG/CR-6850 had negligible impact on the LAR results. This result was supported in the response to PRA RAI 44.01 (Reference 17), by the explanation that the larger zone of influence due to the greater HRR did not result in additional targets or VFDRs being involved in the fire scenarios. Based on the results of the sensitivity study the NRC staff concludes that the risk results are sufficient for use to support transition. However, since the self-approval acceptance guidelines are small, the NRC Staff concludes that the method for modeling transient fires should be revised to either use the HRR distribution in NUREG/CR-6850 or appropriate justification developed for the use of HRRs different than NUREG/CR-6850 before the quantitative results may be used to support future self-approval. In its response to PRA RAI 84 (Reference 21), the'licensee committed to update the Fire PM model to incorporate the sensitivity results for changing the transient fire HRR to 317 kW, and included the actibn as item 24 in the updated Attachment S of the LAR.

The NRC staff noted that new information indicated that the reduction in hot short probabilities for circuits provided with control power transformers (CPTs) identified in NURElCR-6850 were too high and should be reduced. In the response to PRA RAI12 (Reference 13), -credit for CPTs was removed for a sensitivity a"nalysis and the results showed negligible delta risk increase in CDF and LERF for impacted fire areas. While the sensitivity results show no impact for the transition, the NRC staff does not endorse a factor of 2 credit in modeling the CPTs in the Fire PRA without further justification. Since the self-approval acceptance guidelines are small, the NRC staff concludes that the reduction credit for CPTs should be removed for use of the Fire PRA for self-approval. In its response to PRA RAI 84 (Reference 21), the licensee.

- 61 committed to update the Fire PRA model to incorporate the sensitivity results for removing the CPT factor of 2, and included this action as item 23 in the updated Attachment S of the LAR.

The original electrical cabinet factor approach noted in F&Os 4-22, 4-23, and 4-25 was replaced with the Hughes Generic Fire Modeling to address multi-point treatment, fire growth, and fire severity. The NRC staff does not endorse the electrical cabinet factor approach (Reference 63).

The NRC staff reviewed the DAEC fire modeling approach as discussed in Sections 3.4.2.3, 3.8.3, and Attachments A and B. In the response to PRA RAI 84 (Reference 21), this fire modeling approach was confirmed to be in the Fire PRA post-transition model. Based on the replacement of the electrical cabinet factor method with this fire modeling approach, the NRC staff considers this issue resolved.

The use of unsupported severity factors had also been noted. Severity factors are probabilities of fire spreading beyond the initial ignition source (e.g., beyond an electrical cabinet to other nearby combustibles) via fire effects (heat, flame, radiation, etc.) if no action is taken to retard the spread (e.g., without any suppression attempt). Such factors often consider the physics of fire behavior result from fire phenomenological modeling or statistical analysis of historical/experimental experience An unsupported severity factor had been -used for transient scenarios as noted in F&O 4-32.

This was removed from the Fire PRA, and, as noted in response to PRA RAI 21 (Reference 12),

the transient fire severity factor was treated consistently with NUREG/CR-6850, and, therefore, acceptable to the NRC staff. In additio'h; a severity factor had been applied to credited doors for the EDG rooms. In response to PRA RAI 07 (Reference 13), the severity factor was removed, consistent with the use of a barrier failure probability from NUREG/CR-6850, and, instead included credit for the automatic suppression system. The NRC staff, therefore, considers:

these issues resolved.

Differences from NUREG/CR-6850 guidance were also noted in the cable spreading room analysis.Section V.2.1 of the LAR and F&O 5-29 identified the use of a hot work pre-initiator factor to reduce the hot work fire frequency. Also, the NRC staff noted during the audit that a different transient fire influencing factor had been assumed for maintenance for evaluating the transient frequency. In the licensee's response to PRA RAI 14 (Reference 13), which requested an analysis with no deviations from NUREG/CR-6850, the NRC staff further noted that prompt suppression of transient fires had also been credited, and identified other analysis assumptions that required justification. In response to PRA RAI 14.01 (Reference 17), which requested additional justification for the analysis, DAEC performed a re-evaluation of the cable spreading ropm (CSR) analysis with these differences removed, and its results showed negligible delta risk increase in CDF and LERF. The response to PRA RAI 84 confirmed that acceptable methods were applied for the CSR analysis, and are, therefore, acceptable to the NRC staff.

The NRC staff noted that the probability of failure for multi-element rated barriers used in the multi-compartment analysis (MCA) was for the most bounding element in the barrier and not the sum of the barrier failure probability which is logically the proper value. In response to PRA RAI 5 (Reference 13), the probability of barrier failure was revised as the sum of all barrier elements, consistent with NUREG/CR-6850, and, therefore, acceptable to the NRC staff. The response to PRA RAI 84 (Reference 21) confirms that the sum of the barrier elements is in the Fire PRA MCA.

- 62 In PRA RAI 8.01 (Reference 64), the NRC staff noted that the steps for screening MCAs in NUREG/CR-68S0 had not been completely followed. The response to the PRA RAI (Reference 18) was to re-evaluate the screening process identified in NUREG/CR-68S0 for.

MCAs. The updated MCA resulted in the scenarios being screened having a total screening COF less than ten percent of the total estimated plant fire core damage frequency (CO F),

consistent with SR QNS-C1 from the ASME/ANS PRA standard (Reference S9) and as clarified in RG 1.200, Revision 2 (Reference 26).Therefore, the NRC staff considers this issue resolved.

The NRC staff also noted in PRA RAI 72 (Reference 64), that NUREG/CR-68S0 guidance provides a minimum non-suppression probability that should be used but that OAEC had used..

smaller values. In the response to the PRA RAI (Reference 18), OAEC used the minimum non suppression probability in instances in which a smaller non-suppression probability had been applied. This evaluation led to the insight that the automatic suppression system in physical analysis unit (PAU) 030 should be required for risk considerations, whereas, previously, it was only required for defense-in-depth. The response to PRA RAI 84 (Reference 21) confirmed that the Fire PRA uses the minimum non-suppression probability according to NUREG/CR-68S0.

Therefore, the NRC staff considers this issue resolved.

The NRC staff also requested additional information in PRA RAI15 (Reference 41), related to a method for main control room (MCR) panel fires for cabinets separated by a single wall. The response to the RAI stated that the fire PRA used a value of 1 E-3 for the probability that an exposing control room cabinet fire results in fire damage to an adjacent cabinet when separated by a single wall. In response to subsequent PRA RAI 3S.01 (Reference 18) the licensee reported that the MCR analysis was reevaluated using the guidance of NURE!3/CR-68S0 which results in a value of 4E-S instead of 1 E-3. The NRC staff concludes that the single wall separated cabinets have been appropriately evaluated because the accepted guidance of NUREG/CR-68S0 was used.

The MCR analysis also took credit for a location-specific conditional probability applied to the transient frequency. Justification was provided in the response to PRA RAI 71 (Reference 17) for this factor; however, the NRC staff did not endorse this conditional probability because the justification was simply a postulation that transient material was less likely to be stored in a given location. This factor was removed from the Fire PRA in the integrated analysis reported in the response to PRA RAI 82 (Reference 19).

Regarding electronics as discussed in NUREG/CR-6850 Section H.2, NRC PRA RA169, (Reference 64), requested a revised analysis using lower damage criteria for electronics or a demonstration that the use of cable damage criteria in cabinets is acceptable. In response to PRA RAI 69 (Reference 18), the licensee explained that tlie lower temperature threshold of 65 degrees C would not be reached before MCR abandonment occurred. In addition, in response to PRA RAI 82, which requested additional information for such electronics outside the MCR, the licensee updated the Fire PRA to account for component failures, using NUREG/CR-68S0 Section H.2 in determining the threshold damage temperature. Any component that could contain semi-conductors, resistors, capacitors (electrolytic or ceramic), or similar solid-state components was included. Based on the use of the electronics temperature threshold given in NUREG/CR-68S0 Section H.2 in place of the use of the cable damage criteria, the NRC staff considers this issue resolved.

In response to PRA RAI 22 (Reference 12), on emergency diesel generator (EDG) fires, DAEC evaluated EDG fire events to determine if they were potentially challenging and should be included in a Bayesian update of the generic 'fire frequencies. The licensee's review determined that the events were not potentially challenging in accordance with NUREG/CR-6850 guidance, and that the current Bayesian update was appropriate. Therefore, the NRC staff considers this issue resolved.

The NRC staff requested clarification on the turbine generator fire ignition frequency in PRA RAI 11.01 (Reference 64). The response (Reference 18) provided a re-analysis of the catastrophic turbine generator ignition frequency with no additional credit taken for manual suppression. In response to PRA RAI 84 (Reference 21), the licensee confirmed that the catastrophic turbine generator fire modeling in the Fire PRA was updated to be consistent with NUREG/CR-6850 guidance. Therefore, the NRC staff concludes that this issue is resolved.

The human reliability analysis (HRA) modeling approach relied on the Electric Power Research Institute's HRA approach which is discussed in NUREG-1921 (Reference 65). As explained in the responses to PRA RAI 39 (Reference 12), and PRA RAI 39.01 (Reference 17), the cognitive error assessments used both the Caused-Based Decision Tree Method (CBDTM) and AC,cident Sequence Evaluation Program (ASEP) method. The approach taken was to add the results of the CBDTM and ASEP evaluations for recovery times less than one hour, which is the assumed duration of fires. While ASEP is a method which does not consider performance shaping factors directly, CBDTM does. The NRC staff concludes that this combined approach is an acceptable method since the sum of the two methods is conservative with respect to using the CBDT method only, DAEC proposed in the LAR to transition with a Capability Category I (CCI) assignment for using generic unreliability and unavailability with regard to suppression systems credited in the Fire PRA. The responses to PRA RAI 10 (Reference 12), and PRA RAI 10.1 (Reference 17),

provide the justification. DAEC only credits suppression systems for multi-compartment interactions which are not a significant risk contributor to total plant risk. Therefore, the total plant risk is not sensitive to the uncertainty in suppression system unreliability and unavailability.

Furthermore, TB1 (Turbine Building) is the fire area where suppression systems have been credited for multi-compartment interactions in the fire risk evaluations. The analyses where the systems are credited do not include VFDRs; therefore, the Fire PRA delta risk is not sensitive to the uncertainty in suppression system unreliability and unavailability. Furthermore, credited suppression systems will be evaluated as part of the NFPA'805 monitoring program. Based on risk insights for the suppression systems that these systems are not risk significant for the application, and the monitoring program to trend performance, the NRC staff concludes that CCI is acceptable.

In addition to reviewing F&Os, the NRC staff noted Fire PRA modeling aspects and risk inSights relevant to the application. These are discussed below.

As identified in the response to PRA RA131.01 (Reference 17), spurious operation combinations are associated with some containment isolation valves (CIVs) in the model, while other CIVs did not include spurious operation in the model because of a low probability of a pipe break and a spurious operation leading to containment isolation failure. SR ES-B5 in the

- 64 ASME/ANS PRA standard (Reference 59) provides criteria by which fire-induced spurious operation may be excluded. The NRC staff notes that the peer review found SR ES-B5 to be met with no F&Os; therefore, the NRC staff concludes that exclusion of spurious operation of these CIVs was consistent with this SR and their spurious operation is not expected to be a significant contributor for their respective system models.

The NRC staff noted during the audit that the EDGs were located in the turbine building (TB),

and requested in PRA RAI 51 (Reference 41) that it be verified that a TB fire could not deterministically result in a loss of offsite power (LOOP). This is an important Fire PRA consideration since the catastrophic TB fire frequency is not low. In the response to PRA RAI 51 (Reference 13), DAEC concluded that offsite power would be available based on deterministic analysis. A probabilistic conditional LOOP, given TB fires, however, is included in the Fire PRA model logic. Based on the response to the PRA RAI, the NRC staff concludes the Fire PRA model appropriately considered TB fires resulting in a LOOP. '

In addition, dominant fire scenarios identified in the LAR Attachment W result in a loss of high pressure injection systems; therefore, it is important to depressurize to use the low pressure injection systems for inventory control. The response to PRA RAI 43.01 (Reference 17),notes that a postulated fire in the relevant fire areas, i.e., either of the essential switchgear rooms, does not result in failure of the depressurization function. The depressurization model includes relief valve failures, DC power failures, and manual and automatic initiation. Furthermore, the MCA for the essentiaL switchgear rooms resulted in these interactions being screened based on CDF. In addition, the NRC staff's review noted that there is no VFDR-associated risk with the depressurization function in the essential switchgear rooms, or with HPCI and RCIC in these fire scenarios either, as noted in the response to PRA RAI 43 (Reference 13). Based on these RAI responses, the NRC staff concludes that the Fire PRA appropriately considered the depressurization function in the reported dominant sequences.

According to Attachment V of the LAR, F&O 6-3, spatial separation is credited as a partitioning element in the Fire PRA. Justification for the use of spatial separation was provided in response to PRA RAI13 (Reference 12), and relies upon fire modeling. Spatial separation is an important Fire PRA model consideration for FREs of delta CDF and delta LERF since some fire areas where it is credited contain VFDRs which are separation issues. The NRC staff's fire modeling review and findings are discussed in SE Sections 3.4.2.3, 3.8.3, and Attachments A and B.

The NRC staff also reviewed the MCA performed for the Fire PRA. In its response to PRA RAI 40 (Reference 12), DAEC confirmed that the MCA modeling included fire induced failure of the cables which lead to the dominant LOOP sequences. In response to PRA RAI 76 (Reference 18), it is noted that the MCA evaluated increased temperatures for equipment and for adjacent rooms. F&Os 3-8 and 6-4 (Attachment V of the LAR) on passive fire barriers and non-rated fire barriers, respectively, were dispositioned in the responses to PRA RAls 9 and 13 (Reference 12). In its response to PRA RAI 41.01 (Reference 17), on the modeling of a specific missing fire damper between fire areas, DAEC determined that the specific damper noted in the PRA RAI was not an issue for the application, and performed a review of plant drawings and operating practices of dampers between physical analysis units to ensure dampers were appropriately reflected in the Fire PRA. Furthermore, DAEC re-evaluated the MCA screening in response to PRA RAI 8.01 (Reference 18), following NUREG/CR-6850 guidance. Therefore,

- 65 the NRC staff concludes that the responses to these PRA RAls resolve Fire PRA issues identified in the review of the MCA.

DAEC performed Fire PRA qLiantification with the FRANC and XINIT software. FRANC is a software tool used to facilitate a one-time estimate of the core damage and large early release frequencies due to fire when interfaced with a pre-existing "internal events" PRA model. This code replaces the values for pre-existing elements in the PRA model with values that are specific for fire-induced failure, but does not redefine the particular elements, hence its use only as a "one-time" estimate. XINIT is a computer tool that enables multiple computer codes to be integrated automatically when performing a PRA analysis. For example, the results from FRANC could be combined with other tools to produce a nearly completely documented package of all inputs and outputs from a Fire PRA analysis that would meet the documentation requirements from the ASME/ANS-RA-Sa,;,2009 (Reference 59).

The lERF calculation uses the rare event approximation, according to the LAR, which can result in overestimation. The NRC staff considers the CDF and lERF results calculated with FRANC and XINIT to provide the results for the application. DAEC plans to use the FRANC and XINIT software post-transition as noted in the response to PRA RAI 30 (Reference 12).

Fire PRA results for the lAR are provided in Section 3.4.6.

One F&O from the fire PRA peer review, F&O 4-35 on SR FSS-D7, remained at Capability Category I. In response to RAI 10.01 the licensee. c!arified that it did not need to determine whether any fire protection systems have experienced' outlier behavior because such systems are only credited for multi-compartment interactions (which are in general minor contributors to risk) and the analyses where these systems are credited do not include VFDRS and therefore do not contribute to any change in risk. DAEC further clarified that the reliability of these FPP systems will be captured and tracked as part of the NFPA 805 Monitoring Program to be implemented prior to transition. The NRC staff concludes that CCI is sufficient because the reliability of the systems does not contribute to the change in risk in transition, and the post transition NFPA 805 monitoring program requires the reliability to be tracked regardless of the Capability Category assigned.

The Fire PRA has been updated to use methods acceptable to the NRC staff as noted above, or the method had a negligible impact on the change in risk analysis of transition. The results of an integrated study which considered these acceptable methods in response to PRA RAI 82 are discussed in Section 3.4.7 of this SE. The licensee committed to update the Fire PRA model to incorporate only methods acceptable to the NRC Staff before using the PRA to support self approval as Implementation Item 20 in the updated Attachment S (Reference 21).

As a result of its review of the lAR, as supplemented, the NRC staff concludes that the DAEC Fire PRA is sufficiently technically adequate and that its quantitative results, considered together with the sensitivity studies, can be used to demonstrate that the change in risk due to the transition to NFPA-805 meets the acceptance guidelines in RG 1.174 and are acceptable.

3.4.2.3.

Fire Modeling in Support of the Development of the Fire Risk Evaluations (FREs)

The NRC staff performed detailed reviews of the fire modeling used to support the FRE in order to gain further assurance that the methods and approaches used for the application to transitipn

- 66 to NFPA 805 (Reference 7) were technically adequate. NFPA 805 has the following requirements that pertain to fire modeling used in support of the development of the FREs:

NFPA 805, Section 2.4.3.3: On Acceptability The PSA [probabilistic safety assessment] approach, methods, and data shall be acceptable to the AHJ [authority having jurisdiction].

NFPA 805, Section 2.7.3.2, "Verification and Validation":

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

NFPA 805, Section 2.7.3.3, "Limitations of Use":

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

NFPA 805, Section 2. 7.3.4, "Qualification of Users":

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

NFPA 805, Section 2.7.3.5, "Uncertainty Analysis":

An uncertainty analysis shall be performed to provide reasonable assurance that the performance criteria have been met.

The following Sections discuss the results of the NRC staff's reviews of the acceptability of the fire modeling (first requirement). The results of the NRC staffs review of compliance with the remaining requirements are discussed in Sections 3.8.3.2 through 3.8.3.5 of this SE.

3.4.2.3.1 Overview of Fire Models Used to Support the FREs The zone of influence (ZOI) around ignition sources was determined based on tables in the Generic Fire Modeling Treatments (GFMTs) document (Reference 66). The tables in this document provide the horizontal and vertical dimensions of the ZOI for various ignition sources (transient fuel packages, small liquid fuel fires, open cabinets and cable trays) and different types of targets, i.e., thermoplastic and thermoset cables as defined in NUREG/CR-6850 (Reference 62), and Class A combustibles. The GFMTs document also contains a set of tables that are used to determine if and when the hot gas layer (HGL) temperature exceeds the damage threshold of specified targets depending on fire size, room volume, and ventilation

- 67 conditions. The GFMTs document was used as a basis for the scoping or screening evaluation as part of the DAEC fire modeling to support FRE.

During the audit, the NRC staff reviewed supplementary material to the GFMTs document which provides calculated HGL temperatures tables for additional critical temperatures and ignition source heat release rates (HRRs), including some combinations of an ignition source and an intervening combustible. The NRC staff reviewed this material to determine whether the models used were appropriate. The results of the review are discussed below.

The lOI tables in the GFMTs document and its supplementary tables were obtained by using a collection of algebraic models and empirical correlations. The primary algebraic fire models and empirical correlations that were used for this purpose are the following:

Heskestad Flame Height Correlation Heskestad Plume Temperature Correlation Modak's Point Source Radiation Model These algebraic models are described in NUREG-1805, "Fire Dynamics Tools (FDT S):

Quantitative Fire Hazard Analysis Methods for the US Nuclear Regulatory Commission Fire Protection Inspection Program" (Reference 67). Validation and Verification (V&V) of these algebraic models is documented in NUREG-1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications," Volume 3 (Reference 68). The V&V of the fire models that were used to support the DAEC fire PR,A is discussed in Section 3.8.3.2 of this SE.

The Consolidated Model of Fire and Smoke Transport (CFAST) computational fire model, Version 6, was used to generate the HGL tables in the GFMTs document and its supplementary material. The fire PRA used the'se calculations to further screen ignition sources, scenarios, and compartments that would not be expected to generate an HGL, and to identify the ignition sources that have the potential to generate an HGL for further analysis. CF AST was also used for the Main Control Room (MCR) abandonment time calculations. The V&Vof CFAST is documented in NUREG-1824, Volume 5 (Reference 68).

The licensee also identified the use of the following empirical models that are not addressed in NUREG-1824,in the development of the GFMTs document and its supplementary material.

Shokri and Beyler flame radiation model (Reference 69)

Mudan flame radiation model (Reference 70)

Plume heat flux correlation by Wakamatsu et al. (Reference 71)

Yokoi plume centerline temperature correlation (References 72 and 73)

Hydrocarbon spill fire size correlation (Reference 74)

Flame extension correlation (Reference 75)

- 68 Delichatsios line source flame height model (Reference 76)

• Corner flame height correlation (Reference 75)

Kawagoe natural vent flow equation (Reference 77)

• Yuan and Cox line fire flame height and plume temperature correlations (Reference 78)

Lee cable fire model (Reference 79)

Babrauskas method to determine ventilation-limited fire size (Reference 80)

The V&V of these models is discussed in Section 3.8.3.2 of this SE.

The licensee's ZOI approach was used as a screening tool to distinguish between fire scenarios that required further evaluation and those that did not. The licensee stated that qualified personnel performed a plant walk-down to identify: ignition sources, surrounding targets, and safety related SSCs and applied the GFMTs approach to assess whether the SSCs were within the ZOI of a fire scenario. Based on the fire hazard present in the fire areas, these generalized ZOls were used to screen from further consideration those DAEC-specific ignition sources that did not adversely affect the o*peration of credited SSCs or targets, following a fire. The licensee's screening was based on the 98th percentile HRR from the NUREG/CR-6850 methodology.

3.4.2.3.2 RAls Pertaining to Fire Modeling in Support of the DAEC Fire PRA By letters dated January 31,2012 (Reference 41) and November 8,2012 (Reference 45), the NRC staff sought additional information concerning the fire modeling conducted to support the fire PRA. By letters dated April 23, 2012 (Reference 12), May 23,2012 (Reference 13), and January 11,2013 (Reference 16), the licensee responded to these RAls. In addition, by letter dated June 25, 2012 (Reference 81), NRC staff transmitted a list of questions that resulted from a second site audit. The purpose of the second audit was to obtain more detailed information for specific fire areas where fire modeling was performed. By letter dated October 15, 2012 (Reference. 15), the licensee provided a response to these questions. The second audit questions were subsequently transmitted to the licensee in the form of RAls by letter dated October 26,2012 (Reference 82). The following paragraphs describe selected RAI responses*

related to the acceptability of the fire models used.

The NRC staff issued fire modeling (FM) RAI 01 (a) (Reference 41), to ask the licensee to explain how wall and corner effects were accounted for in the MCR abandonment time study and the ZOI determination throughout the plant; and to describe the data collection method for specific ignition sources identified as being in close proximity to walls and/or corners.

In its response (Reference 13), the licensee performed a sensitivity study to determine the effect of wall and corner fires on MCR abandonment. The analysis shows that transient fires in a corner may significantly reduce the MCR abandonment time with the HVAC system in

- 69 normal or purge mode. However, the risk evaluation (core damage frequency (CDF>>) for a 98th percentile transient fire in a corner was estimated to be negligible (1 E-8/year).

Furthermore, as far as the use of the GFMTs is concerned, the licensee stated that fire location effects were accounted for if the fuel package (transient, small fuel spill or open cabinet) was within 2 ft of a wall or corner. The location of each fuel package was noted on the walk-down sheets. To determine the GFMTs ZOI for wall fires; the HRR of the fuel package was doubled and the center of the package was moved closer to the wall. For corner fires; the HRR was quadrupled and the center of the package was moved closer to the corner.

Based on the results of the sensitivity analysis, the NRC staff concludes that the use of the MCR abandonment time study without consideration of wall and corner fires to be acceptable. In addition, NRC staff concludes that the licensee's approach to account for wall and corner effects in theZOI calculations is acceptable.

• The NRC staff issued FM RAI 01 (b) (Reference 41), to determine if and how the modification to the critical heat flux for a target that is immersed in a thermal plume, as described in Section 2.4 of the GFMTs document, was used in the analyses to support the transition to NFPA 805.

I n its response (Reference 13), the licensee explained that based on Section 2.4 calculations, the following damage thresholds were used for cable targets in the determination of the horizontal ZOI dimensions: (1) 11.4 kW/m2 for ambient temperatures of 80 0 C or less, (2) 9.0 kW/m2 for ambient temperatures between 80 0 C and 120 0 C, and (3) 5.7 kW/m2 for ambient temperatures above 120 0 C.

NRC staff concludes that the licensee's approach to account for the additional convective heat transfer in the determination of the horizontal ZOI for cable targets immersed in a heated environment is acceptable because this approach is considered to be conservative.

• The NRC staff issued FM RAI 01 (c) (Reference 41), to obtain clarification of the HGL refinements that were applied in the essential switchgear rooms.

In its response (Reference 13), the licensee performed additional walk-downs of the essential switchgear rooms; 10Eand 10F, to verify location of ignition sources, ventilation configuration, and secondary combustibles. They also explained the approach that was used to determine, based on the tables in the GFMTs and its supplementary material, if and when an HGL is expected to develop. The results of this HGL analysis for all postulated scenarios are summarized in two tables, one for each switchgear room.

The results of the HGL analysis for the essential switchgear rooms indicate that the licensee's approach is conservative. Therefore, the NRC staff concludes that this approach is acceptable.

• The NRC staff issued FM RAI 02(Reference 41) to request that the licensee provide substantiation for the exclusive use of the GFMTs ZOI tables for cable 'targets with thermoset damage thresholds as defined in NUREG/CR-6850, Volume 2, (i.e., 330°C and

III

- 70 11 kW/m2.) It was noted that the GFMTs documents referred to these cable targets as "IEEE-383 Qualified".

At the time the response to FM RAI 02 (Reference 41) was submitted, the licensee had retrieved information on cable material and fire test qualification of 13,300 of the 14,000 cables in the plant. Based on this information the licensee identified 459 cables with thermoplastic insulation jacket material. Plant walk-downs identified no additional targets in the fire areas where thermoplastic cables are present. Since no unqualified thermoplastic cables were found in fire areas modeled in the FRE per NUREG/CR-6850, Volume 2, self ignited cable fire scenarios do not need to be postulated.

NRC staff issued a follow-up request, FM RAI 02.01 (Reference 45), to inquire about the 700 cables that had not been characterized at the time RAI 02 response was sent.

The licensee was not able to retrieve information for all 700 cables. The unidentified cables were therefore assumed to be,unqualified thermoplastic. Self-ignited cable fire scenarios were postulated in plant locations included in the Fire PRA where unidentified cables are present. Additional walk-downs were performed to identify new cable targets or changes in the fire scenario input due to the presence of identified or assumed thermoplastic cables.

Inclusion of fire scenarios that involve unqualified and/or thermoplastic cables in the fire PRA results in an increase of the plant CDF and LERF of 1.6% and 2.5%, respectively.

Given the small increase in plant risk due to the presence of some unqualified and/or thermoplastic cables, and that the licensee confirmed in response to PRA RAI 84 (Reference 21) that the Fire PRA includes the new self-ignited cable fire scenarios and the change in input for fire scenarios with thermoplastic cable targets, the NRC staff concludes that the current analysis, which is based on the assumption that all cables in the plant are thermoset, has been updated and is sufficient to support the requested licensing amendment.

The NRC staff issued FM RAI 07(a) (Reference 45), to ask the licensee to justify the assumptions that were made concerning fire propagation in cable trays in the development of the HGL tables in the GFMTs supplementary m~terial for fires that involve intervening combustibles (cable trays).

In its response (Reference 15), the licensee determined that the approach for addressing secondary combustibles described in the GFMTs supplementary material may yield a HRR that is lower than the value recommended in NUREG/CR-6850 (Reference 62). As a result, the licensee decided to replace those HGL tables with new tables that are based on the cable tray fire propagation model described in NUREG/CR-6850, Volume 2 and a HRR per unit area of 250 kW/m2 as recommended in NUREG/CR-7010 (Reference 83). The postulated fire scenarios in which secondary combustibles are assumed to ignite were reviewed using the revised HGL tables.

Since the licensee's review indicates that the original HGL analysis performed in support of the FRE is conservative compared to a revised analysis based on the cable tray fire propagation model in NUREG/CR-6850, Volume 2, NRC'staff concludes that the current analysis is acceptable.

- 71

• The NRC staff issued FM RAI 07(b) (Reference 45), to ask the licensee to provide a justification for not using tile method described in Section R.4.2 of NUREG/CR-6850, Volume 2 to model fire propagation in cable trays, and to confirm that the GFMTs lOI for cable tray fires are acceptable.

In its response (Reference 15), the licensee developed a revised set of lOI tables for fires that involve cable trays based on the fire propagation model described in NUREG/CR-6850, Volume 2 and as recommended in NUREG/CR-7010, a HRR of 250 kW/m2, Each scenario, in which secondary combustibles are postulated to ignite, was re-assessed using the revised set of lOI tables and did not result in additional risk-significant targets, Since the licensee's re-assessment indicates that the revised lOI analysis was based on the cable tray fire propagation model in NUREG/CR-6850, Volume 2 did not result in new targets with risk-significant cables,the NRC staff concludes that the current lOI analysis for fires that involve cable trays is sufficient to support the requested licensing amendment.

3.4.2.3.3

. Conclusion for Section 3.4.2.3 Based on the licensee's description in the LAR, as supplemented, of the DAEC process for perfo'rming FM in support of the FREs, the NRC staff concludes that the licensee's approach for meeting the requirements of NFPA 805, Section 2.4.3.3 is acceptable.

3.4.2.4 Conclusions Regarding Fire PRA Quality

('

Based on NUREG-0800, Section 19.2, Section 111.2.2.4.1, ~ummarizing the NRC Staff's review of PRA Quality required for an application, the NRC staff concludes that the licensee's PRA satisfies the guidance in RG 1.174, Section 2.3, and RG 1.205, Section 4.3 regarding the technical adequacy of the PRA used to support risk assessment to support transition to NFPA 805.

The NRC staff concludes that the PRA approach, methods and data are acceptable and therefore that Section 2.4.3.3 of NFPA-805 is satisfied for the request to transition to NFPA-805.

The NRC staff based this conclusion on the findings that: (1) the PRA model for DAEC meets the criteria in that it adequately represents the current, as built, as operated configuration, and is therefore capable of being adapted to model both the post-transition and compliant plant as needed; (2) the PRA models conform sufficiently to the applicable industry PRA standards for internal events and fires at an appropriate Capability Category, considering the acceptable disposition of the peer review and NRC staff review findings; and (3) the fire modeling used to support the development of the DAEC Fire PRA has been confirmed as appropriate and acceptable.

Prior to using the Fire PRA results to support risk-informed self-approval of changes to the FPP the NRC staff concludes that the following must be completed since the self-approval acceptance guidelines are much smaller than the transition acceptance guidelines.

• The NRC staff concludes that a number of changes identified in the response to PRA RAI 84 (Reference 21) replacing unacceptable methods with acceptable methods have

(

- 72 been made to the fire PRA although not all changes have been completed and/or fully documented in the plant documents. This will be resolved upon completion of Implementation of the PRA update described in Item 20 of the Updated Table S-2.of the LAR.

NRC staff concludes th"at the credit for CPTs should be removed because new information as noted in PRA RAI 12 (Reference 13) has determined the credit is not justified. This will be resolved upon completion of Implementation of the PRA update described in Item 23 of the updated Table S-2 of the LAR.

NRC staff concludes that the method for modeling transient fires should be revised to either use the HRR distribution in NUREG/CR-6850 or appropriate justification developed for the use of HRRs different than NUREG/CR-6850. This will be resolved upon completion of Implementation Item 24 of the updated Table S-2 of the LAR.

In addition, Implementation Item 22 of the updated Table S-2, states that the licensee will update the PRA to model the as-built TB ESW pump B modification upon completion of the modification. If the revised model shows that the change in risk-guidelines for self-approved changes, are exceeded, then the results will be entered into the corrective action program to determine the cause and corrective actions.

Finally, based on the licensee's administrative controls to maintain the PRA models current and assure continued quality, using only qualified staff and contractors (as described in Section 3.8.3 of this safety evaluation), the NRC staff concludes that the PRA maintenance process can assure that the quality of the DAEC PRA is sufficient to support self-approval of future risk-informed changes to the FPP under the NFPA 805 license condition following completi!,>n of all implementation items described in the updated Table S-2.

3.4.3 Fire Risk Evaluations For those fire areas for which the licensee used a performance-based approach to meet the nuclear safety performance criteria, the licensee used fire risk evaluations in accordance with NFPA 805 Section 4.2.4.2 to demonstrate the acceptability of the plant configuration. In accordance with the guidance in RG 1.205, Section C,2.2.4, "Risk Evaluations," the licensee used a risk-informed approach to justify acceptable alternative to comply with NFPA 805 deterministic criteria. The NRC staff reviewed the following information during its evaluation of DAEC's fire risk evaluations: LAR Section 4.5.2, "Performance Based Approaches," LAR Attachment C, "NEI 04-02 Table B Fire Area Transition," and LAR Attachment W, "Fire PRA Risk Insights," as well as associated supplemental information.

Plant configurations that did not meet the deterministic requirements of NFPA-805, Section 4.2.3.1 were considered VFDRs. VFDRs that will be brought into deterministic compliance through plant modifications need no risk evaluation. The licensee identified 62 VFDRs in the LAR Attachment C, NEI 04-02 Table B Fire Area Transition, that it does not intend to bring into deterministic compliance under NFPA 805. For these VFDRs the licensee performed '

evaluations using the risk-informed approach, in accordance with NFPA 805, Section 4.2.4.2, to address FPP non-compliances and demonstrate that the VFDRs are acceptable.

- 73 All of the VFDRs identified at DAEC by the licensee were categorized as separation issues.

Generally, VFDRs were identified for the following types of plant configurations: (1) inadequate separation resulting in fire-induced damage of process equipment, or associated power and/or control cables, required for the identified success path; (2) inadequate separation resulting in fire-induced spurious operation of equipment that may defeat the identified success path; (3) inadequate separation resulting in fire-induced failure of process monitoring instrumentation, or associated cables, required for the identified success path; and (4) combinations of the above configurations.

In response to PRA RAI 23 (Reference 13), the licensee explained that ERFBS is limited to use in fire area RB1 at DAEC but that this fire wrap is not credited in the Fire PRA since fire damage to wrapped cables was postulated consistent to that of exposed cables. The licensee also stated that the embedded cables in fire area TB1 are in a concrete chase, and while the concrete chase may potentially be subject to mechanical damage, the damage is unlikely to diminish the fire rating of the chase to the extent that the protected cables would be damaged by postulated fires. The concrete chase in fire area TB1 is a rated-three hour barrier. Because of the three-hour rating of the concrete chase, the cables embedded in the concrete chase are not the subject of a VFDR and are credited in the Fire PRA as being protected from fire damage. The NRC staff concurs that cables embedded in concrete are unlikely to be damaged by most credible fires and therefore less credit than the full rating of the fire barrier is unnecessary.

The response to PRA RAI 60 (Reference 12), summarizes how a FRE is performed for a VFDR.

Each VFDR is reviewed to ensure that it is adequately reflected in the Fire PRA model. The variant case is with the VFDR present. The compliant case removes the VFDR to represent a deterministically compliant condition. This is accomplished by not failing the basic events that would be failed by the fire in the fire scenarios associated with the VFDR. In the compliant case, the basic events that would be failed by the fire are set to their random failure probabilities. The change in risk is the variant minus the compliant case. In addition, according to the response to PRA RAI 60.01 (Reference 17), no recovery actions were added to the Fire PRA. For fire scenarios which involve more than one VFDR, the delta risk calculation includes synergistic effects from the cables and equipment associated with each VFDR included in the fire scenario.

The planned modification for the TB ESW Pump B will remove a VFDR in fire area TB1, i.e.,

VFDR SSA-TB1-09, in Attachment C of the LAR. This modification is appropriately included in the variant and compliant cases of the.6CDF and.6LERF. The modification will also remove a VFDR in fire area RB4, but according to Table 5 in the response to PRA RI 82 (Reference 19),

there is no impact on fire area RB4.6CDF and.6LERF results.

In its responses to PRA RAls 60.01d (Reference 17), 64 (Reference 16), and 84 (Reference 21), and 14.01(Reference 17), the licensee described how it estimated the risk associated from fires that lead to main control room (MCR) abandonment, and the change in risk ariSing from accepting VFDRs that cause abandonment. When the MCR is abandoned, the facility is shut down using the alternative shutdown capability (ASC). The design of the ASC was evaluated by the NRC and concluded to be in accorda'nce with applicable requirements as indicated in an NRC Safety Evaluation dated January 6, 1983 (Reference 58). Shutdown is accomplished USing designated ASC equipment to accomplish core spray and residual heat

- 74 removal powered by a standby diesel generator. The licensee reported that there are detailed procedures and associated training on shutting down the reactor after MCR abandonment. The licensee reviewed the actions to establish ASC to identify any actions that do not occur at a primary control station (PCS) (i.e., recovery actions in NFPA-805). The licensee identified the recovery actions in Table G-1 of the LAR.

The licensee described its methodology to evaluate risk from MCR abandonment due to MCR habitability in the response to PRA RAI 64 and PRA RAI 84. The methodology identified three categories of fire induced scenarios characterized by three different levels of challenges to the operators and therefore three different probabilities of failure to successfully shutdown the plant.

In the response PRA RAI 64, the licensee provided the summaries of the characteristic of each category. The simplest challenge occurs when a fire dpes not impact ASC equipment or the equipment is available upon transfer at a primary control station (PCS). The most complex challenge is applied when a single or multiple hot short-induced spurious operations (MSO) impacts ASC equipment and is considered not recoverable by ASC based on the current design and procedures. The intermediate challenge is applied when a fault impacts ASC equipment and is not recoverable at the PCS based on the current design and procedures. The NRC staff concludes that the use of three categories appropriately characterizes the complexity of the subsequent required operator actions to the extent necessary to differentiate between the failure probabilities.

The licensee's response in PRA RAI 84 stated that fire sequences in which no fire damage occurred to the ASC (the simplest category discussed above) relied on a recovery action in the

,variant case and did not rely on a recovery action in the compliant case. Given this, the recovery action should contribute to the transition change in risk. However, the licensee stated that hardware failure probabilities bound recovery action failure probabilities and therefore estimated the change in risk for the simplest category to be zero. The NRC staff did not agree with this assumption but, as summarized in Section 3.4.7 of this SE, observed that total risk from MCR abandonment (i.e., without subtracting the compliar;lt case risk) when added to the other changes in risk from transition still results in a transition change in risk less than the RG 1.174 guidelines. Therefore accepting the licensee's assumption is not necessary for the NRC staff to conclude that the change in risk from transition is acceptable.

The NRC staff concludes that the licensee's methods for calculating the change in risk associated with VFDRs are acceptable because they are consistent with RG 1.205, Section 2.2.4.1, and FAQ 08-0054 (Reference 36). The NRC staff further concludes that the results of these calculations for each fire area, which are summarized in Table 3.4.6-2 below, demonstrate that the difference between the risk associated with implementation of the deterministic requirements and that of the VFDRs meets the risk acceptance criteria described in NFPA 805, Section 2.4.4.1.

- 75 3.4.4 Additional Risk Presented by Recovery Actions The NRC staff reviewed LAR Attachment C, "NEI 04-02 Table B Fire Area Transition,"

Attachment G, "Recovery Actions Transition," and Attachment K, "Existing Licensing Action Transition," during its evaluation of the additional risk presented by the NFPA 805 recovery actions at DAEC. Section 3.2.5 of this SE describes the identification and evaluation of recovery actions.

DAEC used the guidance in RG 1.205, Revision 1 (Reference 7) for addressing recovery actions. This included consideration of the definition of PCS and recovery action, as clarified in the RG 1.205, Revision 1. Accordingly, any actions required to transfer control to, or operate equipment from, the PCS, while required as part of the RI/PB FPP, were not considered recovery actions per the RG 1.205 guidance and in accordance with NFPA 805. Alternatively, any operator manual actions required to be performed outside the control room and not at the PCS were considered recovery actions.

DAEC identified the RAs in the LAR Attachment G, Table G-1, and noted that all of the RAs in Table G-1 are required for defense-in-depth but are not required for risk reduction and are not modeled in the Fire PRA. The DID RAs are new and are not previously approved, according to the PRA RAI 52 (Reference 12) response. The RAs are related to fire scenarios requiring MCR abandonment associated with Fire Area CB1. The NRC staff's evaluation of the fire risk evaluations is discussed in Section 3.4.3. Other fire areas with VFDRs similar to those in CB1 did not result in the need for MCR abandonment. RAs.are not assigned to those VFDRs so that the additional risk of recovery actions for other fire areas is not applicable as noted in the LAR, Attachment W. The total additional risk from recovery actions in fire area CB1 in the Table W-3 replacement provided in response to PRA RAI 82, is 3.5E-8/yr for CDF and 3.5E-8/yr for LERF.

The NRC staff's evaluation of the method to evaluate the additional risk of RAs is discussed in Section 3.4.3 of this SE.

The NRC staff concludes that the licensee's approach for calculating the additional risk of RAs is acceptable because it is consistent with RG 1.205, Section 2.2.4.1, and FAQ 07-0030 (Reference 31), with the exception noted and evaluated in Section 3.4.3 of this SE. This is related to MCR abandonment sequences in which ASC equipment is not fire-impacted and RAs are credited in the variant case and not in the compliant case. The results of the delta risk calculations for each fire area, and the total, are summarized in Table 3.4.6-2 of this SE. As discussed in Sections 3.4.6 and 3.4.7 of this SE, the NRC staff concludes that these results demonstrate that the total risk of transition, of which part is attributable to the risk of recovery actions, is less than the risk acceptance guidelines in RG 1.174 and therefore the additional risk associated with recovery actions is acceptable.

3.4.5 Risk-Informed or Performance-Based Alternatives to Compliance with NFPA 805 The licensee did not use any risk-informed or performance-based alternatives to compliance with NFPA 805.

- 76 3.4.6 Cumulative Risk and Combined Changes DAEC identified a planned NFPA 805 transition modification that removes VFDRs. The licensee did not propose any modifications that were not required to remove VFDRS and therefore, there is no combined change evaluation.

The total CDF and LERF are estimated by adding the risk assessment results for internal events and fire. For seismic events, DAEC had performed a Seismic Margins Assess'ment. CDF and LERF for other hazards were not part of the LAR since RG 1.174 does not require a total CDF and LERF when the ACDF and ALERF are less than 1 E-6/yr and 1 E-7/yr, respectively. The CDF and LERF results are summarized in Table 3.4.6-1.

Table 3.4.6-1: CDF and LERF for DAEC After Transition to NFPA 805 Hazard Grol:Jp Internal Events Fires Seismic (1) These frequencies are reported in Table 1 of the PRA RAI 82 response, and are included in this table since the NUREG/CR-6850 fire ignition frequencies are included in the Fire PRA model. The LERF is conservative as discussed below.

(2) DAEC performed a Seismic Margins Assessment The licensee also provided the ACDF and ALERF estimated for each fire area at DAEC that is not deterministically compliant, in accordance with NFPA 805, Section 4.2.3, "Deterministic Approach." The risk estimates for these fire areas result from the completed and planned modifications and administrative controls that will be implemented as part of the transition to NFPA 805 at DAEC, as well as recovery actions, to reduce VFDR risk. The ACDF and ALERF results by fire area are summarized in Table 3.4.6-2.

Table 3.4.6-2: ACDF and ALERF for DAEC After Transition to NFPA 805 Fire Area ACDF (lyear)

ALERF (lyear)

BA N/A N/A CB1 3.55E-8 3.53E-8 CB2 epsilon epsilon CB3 1.82E-8 1.67E-8 CB4 1.72E-10 2.59E-12 DRY

)

N/A N/A EX1 N/A N/A IS1 N/A N/A

- 77 IS2 N/A N/A PH1 N/A N/A PH2 N/A N/A RB1 7.44E-9 1.07E-9 RB3 6.41 E-8 2.83E-8 RB4 N/A N/A TB1 t.83E-10 4.91E-11 TOTAL 1.26E-7 Each of the individual fire area changes in risk for CDF and LERF fall into Region III of the RG 1.174 acceptance guidelines. From Table 3.4.6-1, the total CDF is less than 1E-4/yr, and the total LERF is close to 1 E-S/yr but not significantly above it. The sensitivity analysis discussed in Section 3.4.7 below shows that this LERF is conservative since use of the EPRI101673S fire bin ignition frequencies as allowed in FAQ 08-0048 (Reference 3S),

decreases the LERF. The decrease provided from this frequency consideration is from 1.6E-S/yr (which uses the Bayesian updated NUREG/CR-68S0 generic fire frequencies), to 1.1 E-S/yr, according to the updated Attachment W tables from the PRA RAI 82 (Reference 19) response. Therefore, the quantified LERF would be approximately 1.2 E-S/yr by summing this LERF with the internal events PRA LERF. However, based on sensitivity studies discussed in Section 3.4.7 below, the NRC staff cannot definitively conclude the LERF is less than 1 E-S/yr.

Based on the results of the licensee's fire risk assessments, as summarized above, the risk increase for each fire area associated with transition to NFPA 80S at DAEC, as well as the cumulative change in risk for all fire areas subject to a performance-based approach, is within the RG 1.174 risk acceptance guidelines of 1 E-S/yr f1CDF and 1 E-7/yr f1LERF. The NRC staff concludes that the RG 1.174 risk acceptance guidelines of 1 E-7/yr f1LERF apply since the total LERF is slightly greater than 1 E-S/yr.

Therefore, the NRC staff concludes that the risk associated with the proposed alternatives to compliance with the deterministic criteria of NFPA 80S is acceptable for the purpose of this application, in accordance with NFPA 80S, Section 2.4.4.1. Additionally, the NRC staff concludes that the licensee has satisfied RG 1.174, Section 2.4, and NUREG-0800, Section 19.2 regarding acceptable risk.

3.4.7 Uncertainty and Sensitivity Analyses The licensee evaluated key sources of uncertainty and sensitivity in response to PRA RAI 27 (Reference 13). The assessment noted one sensitivity analysis which resulted in a significant decrease in CDF and LERF. The DAEC Fire PRA uses the NUREG/CR-68S0 generic fire ignition frequencies and applied a Bayesian update process for events after the year 2000. The sensitivity analysis used the EPRI101673S fire bin ignition frequencies noted in FAQ 08-0048.

The Attachment W results, provided in response to PRA RAI 82 (Reference 19), show that the f1CDF and f1LERF remain in Region III of RG 1.174 with the use of the NUREG/CR-68S0 generic fire ignition frequencies.

- 78 Other sources of uncertainty considered in the RAI response were not found to result in significant impacts on CDF or LERF. One reason for this is that LOOP is a significant contributor to fire-related risk, and there are fire scenarios which result in LOOP that are not recoverable. DAEC's feasibility analysis as noted in F&O 2-13 of Table V-3 of the LAR determined that recovering offsite power in such scenarios was not feasible. Therefore, the NRC staff notes that this non-recovery of offsite power represents realistic modeling rather than conservative modeling. As a result of these dominant fire scenarios, the Fire PRA CDF and LERF are relatively insensitive to a number of sensitivity analyses.

Another important insight provided from the response to PRA RAI 27 (Reference 13), is related to a sensitivity quantification performed for components modeled in the Fire PRA by assumed routing or credited by exclusion, which included a number of containment-related support systems. The sensitivity study found that if these components were assumed to be available for all fire ~cenarios that cable selection for such components would at most result in a small reduction in CDF. Therefore, while the reduction in LERF was not discussed in the response from this sensitivity study, the NRC staff notes that at most a small reduction in LERF would also be expected. Reduction of the LERF may also be expected due to refinements of the cable spreading room analysis and mitigation strategies as noted in the response to PRA RAI 82.

With respect to the MCR abandonment sequences when postulated fire damage would not affect ASC, discussed in Section 3.4.3 of this SE, the response to PRA RAI 84 shows that the total MCR abandonment scenarios CDF is 4.0 E-8/yr and LERF is 4.0E-8/yr. These frequencies include the contribution from MCR abandonment sequences when postulated fire damage would not affect ASC. If these frequencies are substituted in Table 3.4.6-2 for the fire area CB1 as a sensitivity analysis, the total ~CDF and ~LERF remain less than 1 E-S/yr and 1 E-7/yr, respectively, which are the transition acceptance criteria discussed in Section 3.4.6 of this SE.

With respect to differences in methods from NUREG/CR-68S0, an integrated sensitivity analysis had been performed as given in the response to PRA RAI 82. The acceptable methods included in the analysis were discussed above in Section 3.4.2.2. The integrated sensitivity study provided a comparison between two Fire PRA model baselines, the difference being one which included the use of 98 percentile 317kW in place of the 69kW transient HRR and no credit for CPTs in place of a factor of 2, while the other did not include these. The differences between the CDF and LERF and the ~CDF and ~LERF were negligible due to these modeling assumptions. The resolution of these issues is discussed in Section 3.4.2.2 of this SE.

The licensee demonstrated that most assumptions are conservative; thereby assuring that the existing risk analyses reasonably bound any uncertainty. Accordingly, the NRC staff concludes that the licensee's risk evaluations are reasonable and conservative, and not significantly impacted by the specific modeling assumptions made by the licensee.

3.4.8 Conclusion for Section 3.4 Based on the information provided by the licensee in the LAR, as supplemented, regarding the fire risk assessment methods, tools, and assumptions used to support transition to NFPA 80S at DAEC the NRC staff concludes the following:

- 79 The licensee's PRA used to perform the risk assessments in accordance with NFPA 805, Section 2.4.4 (plant change evaluations) and Section 4.2.4.2 (fire risk evaluations), is of sufficient quality to support the application to transition the DAEC FPP to NFPA 805. The NRC staff concludes that the PRA approach, methods, tools and data are acceptable and are in accordance with NFPA 805, Section 2.4.3.3.

• The NRC PRA maintenance process is adequate to support self-approval of future risk informed changes to the FPP following completion of the PRA related implementation items numbers 20, 22, 23 and 24 described in Attachment S of the LAR.

• The transition process included a detailed review of fire protection DID and SM as required by NFPA 805. The NRC staff concludes that the licensee's documentation on DID and SM is acceptable. The licensee's process followed the NRC-endorsed guidance in NEI 04-02, Revision 2, and is consistent with the approved NRC staff guidance in RG 1.205, Revision 1, which provides an acceptable approach for meeting the requirements of 10 CFR 50.48( c).

• The changes in risk (i.e., L1CDF and L1LERF) associated with the proposed alternatives to compliance with the deterministic criteria of NFPA-805 (fire risk evaluations) are acceptable for the purposes of this application, and the licensee has satisfied the guidance contained in RG 1.205, Revision 1, RG 1.174, Section 2.4, and NUREG-0800, Section 19.2, regarding acceptable changes in risk. By meeting the guidance contained in these approved regulatory documents, the changes in risk have been concluded to be acceptable to the NRC staff, and therefore meet the requirements of NFPA 805.

• The risk presented by the use of recovery actions was determined and provided in accordance with the guidance in RG 1.205, Revision 1, and NFPA 805, Section 4.2.4.

The NRC staff concluded that the additional risk associated with each NFPA 805 recovery action is acceptable because the risk for each fire area that relies on a recovery action is below the acceptance guidelines in RG 1.174 and therefore meets the acceptance criteria in RG 1.205, Revision 1.

• The licensee did not use any risk-informed or performance-based alternatives to compliance to NFPA 805 which fall under the requirements of 10 CFR 50.48(c)(4).

3.5 Nuclear Safety Capability Assessment Results NFPA 805 (Reference 7), Section 2.2.3, "Evaluating Performance Criteria," states the following:

To determine whether plant design will satisfy the appropriate performance criteria, an analysis shall be performed on a fire area basis, given the potential fire exposures and damage thresholds, using either a deterministic or performance-based approach.

NFPA 805, Section 2.2.4, "Performance Criteria," states the following:

- 80 The performance criteria for nuclear safety, radioactive release, life safety, and property damage/business interruption covered by this standard are listed in Section 1.5 and shall be examined on a fire area basis.

NFPA 805, Section 2.2.7, "Existing Engineering Equivalency Evaluations," states:

When applying a deterministic approach, the user shall be permitted to demonstrate compliance with specific deterministic fire protection design requirements in Chapter 4 for existing configurations with an engineering equivalency evaluation. These existing engineering evaluations shall clearly demonstrate an equivalent level of fire protection compared to the deterministic requirements.

3.5.1 Nuclear Safety Capability Assessment Results by Fire Area NFPA 805, Section 2.4.2, "Nuclear Safety Capability Assessment (NSCA)," states the following:

The purpose of this section is to define the methodology for performing an NSCA. The following steps shall be performed:

(1) Selection of systems and equipment and their interrelationships necessary to achieve the nuclear safety performance criteria in Chapter 1 (2) Selection of cables necessary to achieve the NSPC in Chapter 1 (3) Identification of the location of nuclear safety equipment and cables (4) Assessment of the ability to achieve the NSPC given a fire in each fire area This section of the SE addresses the last topic regarding the ability of each fire area to meet the NSPC of NFPA 805. Section 3.2.1 of this SE addresses the first three topics.

NFPA 805, Section 2.4.2.4, "Fire Area Assessment," also states the following:

An engineering analysis shall be performed in accordance with the requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the NSPC of Section 1.5.

In accordance with the above, the process defined in NFPA 805, Chapter 4, provides a framework to select either a deterministic or a PB approach to meet the NSPC. Within each of these approaches, additional requirements and guidance provide the information necessary for the licensee to perform the engineering analyses necessary to determine which fire protection systems and features are required to meet the NSPC of NFPA 805.

- 81 NFPA 805, Section 4.2.2, "Selection of Approach," states the following:

For each fire area either a deterministic or performance-based approach shall be selected in accordance with Figure 4.2.2. Either approach shall be deemed to satisfy the nuclear safety performance criteria. The performance-based approach shall be permitted to use deterministic methods for simplifying assumptions within the fire area.

This section of the SE evaluates the approach used to meet the NSPC on a fire area basis, as well as which fire protection features and systems are required to meet the NSPC.

The NRC staff reviewed LAR Section 4.2.4, "Fire Area Transition," LAR S~ction 4.8.1, "Results of the Fire Area Review," LAR Attachment C, "NEI 04-02 Table B Fire Area Transition," LAR Attachment G, "Recovery Actions Transition," LAR Attachment S, "Plant Modifications and Items to be Completed During Implementation" and LAR Attachment W, "Fire PRA Insights," during its evaluation of the ability of each fire area to meet the NSPC of NFPA 805.

DAEC is a single unit BWR with 15 individual fire areas including the Yard (Buffer Area and EX1) and each fire area is composed of multiple fire zones. Based on the information provided in the LAR, as supplemented, the licensee performed the NSCA on a fire area basis. LAR Attachment C provides the results of these analyses on a fire area basis and also identifies the individual fire zones within the fire areas. The licensee documented the following:

Table 3.5-1 of this SE identifies those fire areas that were analyzed using either the deterministic or PB approach in accordance with NFPA 805 Chapter 4 based on the information provided in LAR Attachment C, Table 8-3, "Fire Area Transition".

Table 3.5-1 Fire Area and Compliance Strategy Summary NFPA 805 Compliance Basis Fire Area Area Description Buffer Areas BA Deterministic CB1 Control Building, CSR, MCR, CR HVAC Rm Performance Based West Essential CB2 Performance Based Battery Rm 1 East Essential CB3 Performance Based Battery Rm Rm Corr. and 250VDC CB4 Performance Based DRY Deterministic EX1 Deterministic

.,S1 Intake Structure - Div. 1 Pump Deterministic IS2 Intake Structure Div. 2 Pump Deterministic PH1 RHRSW/ESW Pump House Div. 2 Deterministic PH2 p House Div. 1 Deterministic

- 82 Fire Area Area Description NFPA 805 Compliance Basis RB1 Reactor Building, 757' Elev. NW, SE, SW corner RCIC, HPCI Rms Performance Based RB3 i Reactor Building: 786' Elev. & above Performance Based RB4 Reactor Building, NE Corner Deterministic TB1 Turbine Building Performance Based LAR Attachment C provides the results of these analyses on a fire area basis. For each fire area, the licensee documented the following:

The approach used in accordance with NFPA 805 (i.e., the deterministic approach in accordance with NFPA 805, Section 4.2.3, or the PB approach in accordance with t-JFPA 805, Section 4.2.4).

The SSCs required to meet the NSPC.

Fire detection and suppression systems required to meet the NSPC.

An evaluation of the effects of fire suppression activities on the ability to achieve the NSPC.

The disposition of each VFDRs using either; modifications (completed or committed) or the performance of a FREs in accordance with NFPA 805, Section 4.2.4.2.

3.5.1.1 Fire Detection and Suppression Systems Required to meet the NSPC A primary purpose of NFPA 805 Chapter 4 is to determine, by analysis, what fire protection features and systems need to be credited to meet the NSPC. Four sections of NFPA 805 Chapter 3 have requirements dependent upon the results of the engineering analyses performed in accordance with NFPA 805 Chapter 4: (1) fire detection systems, in accordance with Section 3.8.2; (2) automatic water-based fire suppression systems, in accordance with Section 3.9.1; (3) gaseous fire suppression systems, in accordance with Section 3.10.1; and (4) passive fire protection features, in accordance with Section 3.11. The features/systems addressed in these sections are only required when the analyses performed in accordance with NFPA 805 Chapter 4 indicate the features and systems are required to meet the NSPC.

The licensee performed a detailed analysis of fire protection features and identified the fire suppression and detection systems required to meet the NSPC for each fire area. LAR Table 4-3, "NFPA 805 Ch 4 Required Fire ProteCtion (FP) Systems/Features," lists the fire areas and identifies if the fire suppression and detection systems installed in these areas are required to meet criteria for separation, DID, risk, licensing actions, or EEEEs.

, The NRC staff reviewed LAR Attachment C for each fire area to ensure fire detection and suppression met the prinCiples of DID in regard to the planned FPP transition to NFPA 805.

Based on the statements provided in LAR Attachment C, as supplemented, the NRC staff concludes that the DAEC treat~ent of this issue is acceptable because, the licensee has

- 83, adequately identified the fire detection and suppression systems required to meet the I\\IFPA 805 NSPC on a fire area basis.

3.5.1.2 Evaluation of Fire Suppression Effects on NSPC Each fire area of LAR Attachment C includes a discussion of how the licensee met the requirement to evaluate the fire suppression effects on the ability to meet the NSPC.

The licensee stated that damage to plant areas and equipment from the accumulation of water discharged from manual and automatic fire protection systems and the discharge of mal1.ual suppression water to adjacent compartments is controlled. Therefore, fire suppression activities will not adversely affect the plant's ability fo achieve the NSPC.

Based on the information provided by the licensee in the LAR, as supplemented, the licensee has evaluated fire suppression effects on meeting the NSPC and determined that fire suppression activities will not adversely affect achievement of the NSPC. The NRC staff has reviewed this information and concludes that the licensee's evaluation of the suppression effects on the NSPC is acceptable.

3.5.1.3 licensing Actions Based on the information provided in the LAR Attachment C, the licensee identified exemptions from deterministic requirements for each fire area that were previously approved by the NRC and will be transitioned with the NFPA 805 FPP. Each of the these exemptions is summarized in LAR Attachment C on a fire area basis and described in further detail in LAR Attachment K, "Existing Licensing Action Transition". The licensee does not have any elements of the current FPP for which NRC clarification is needed. The licensing actions being transitioned are summarized in Table 3.5-2.

Table 3.5-2 Previously Approved Licensing Actions Being Transitioned Licensing Action Applicable Clarification [as applicable]

NRC Staff Description Fire Area Evaluation Exemption #04 RB1 The basis for approval as Based on the (19831219), Appendix RB3 described by the licensee in LAR previous staff R Exemption for Fire Attachment K is the separation approval of this Zone Boundaries and configuration of redundant exemption and the Having Communication cables; the low combustible statement by the Paths with Less than 3 loading; the intervening concrete licensee that the Hour Fire Ratings floor; and the partial suppression basis remains valid, Between Zones system. In Section 4.2.3 of the the NRC staff.

(Equipment Hatch)

LAR the licensee stated that this concludes that the (1II.G.2.a Criteria).

licensing action will be applicability of this (Reference 84) transitioned into the NFPA 805 licensing action is FPP as previously approved acceptable.

(NFPA Section 2.2.7) and that this licensing action is considered compliant under 10CFR50.48(c).

- 84 Licensing Action Clarification [as applicable]

NRC Staff Description Evaluation Exemption #15 CB1 The basis for approval as Based on the (19871014), Appendix CB2 described by the licensee in LAR previous staff R Exemption from the CB3 Attachment K for the exemption approval of this Requirement that CB4 includes that two fire modeling exemption and the Structural Steel IS1 methodologies are employed in statement by the Forming Part of or IS2 the analysis: a fully developed licensee that the Supporting Fire PH1 enclosure fire model is used to basis remains valid, Barriers be Protected to PH2 evaluate the average gas mixture the NRC staff a Fire.

RB1 temperature in the enclosure; concludes that the (Reference 85)

RB3 and the local heating effects on applicability of this RB4 steel members are assessed by licensing action is flame and fire plume acceptable.

impingement calculations. These models formed the basis of the structural steel evaluation. In addition, the licensee stated that if the steel temperature exceeds 11 OO°F within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, then Iowa Electric Light and Power Company (IELP) had committed to protect the steel with 3-hour rated fireproofing, but if the steel temperature does not reach 1100°F, an exemption from the requirements to provide structural steel fireproofing was requested.

In addition, IELP had committed to institute operational procedures to ensure that the combustible load limit assumed by the calculations is not exceeded.

ea Exemption #16 DRY:

The basis for approval as Based on the (19910816), Appendix Drywell described by the licensee in previous staff R Exemption from the Attachment K of the LAR is that approval of this 3-Hour Fire Barrier most of the foam material was exemption and the Requirement for the removed from the expansion gap statement by the Drywell Expansion Gap at DAEC following each concrete licensee that the (1II.G.2.a Criteria).

pour; the only combustible basis remains valid, (Reference 86) material rE?maining in the the NRC staff expansion gap at DAEC is elastic concludes that the polyurethane circumferential applicability of this strips 3 inches thick x 5 inches licensing action is wide on 2-feet centers below acceptable.

elevation 748 feet 9 inches and 3-feet centers above that

- 85 Licensing Action Description Applicable Fire Area Clarification [as applicable]

NRC Staff Evaluation elevation; the strips are manufactured of plastic material that is classed as "self extinguishing" in accordance with American Society for Testing and Materials (ASTM) D 1692; because of the geometry (long narrow circumferential strips separated by 3 feet on centers from below the equator of the bulb) and the self-extinguishing characteristics of the plastic material,any fire that might occur is expected to be limited to the area of ignition and would not spread to other strips; the steel drywell itself will serve as a large heat sink to further assist in cooling and aiding the self-extinguishing characteristics of this material should it become ignited; and maintenance work on containment penetrations is administratively controlled. In addition to fire watches, precautions include filling the annulus space with non combustible material prior to any operations which might produce, hot slag or sparks.

The NRC staff reviewed the exemptions from the pre-NFPA 805 licensing basis identified in Table 3.5-2, including the description of the previously approved exemption from the deterministic requirements, the basis for and continuing validity of the exemption, and the NRC staff's original evaluation or basis for approval of the exemption. The licensee stated in LAR Section 4.2.3, that the review of these existing licensing actions included a determination of the basis of acceptability and that determination of acceptability was still valid.

Based on the NRC staff's review of the licensing actions identified and described in LAR Attachments C and K, the NRC staff concludes that the Licensing Actions are identified by applicable fire area and remain valid to support the proposed license amendment because the licensee used the process described in !\\lEI 04-02 as endorsed by RG 1.205, which requires a determination of the basis of acceptability and a determination that the basis is still valid. Based on the previous staff approval of the exemptions and the statement by the licensee that the basis remains valid as presented in each appropriate fire area, the NRC staff concludes that the

. - 86

'engineering E1valuations being carried forward supporting the NFPA 805 transition, as identified in Table 3.5-2, are acceptable. See Section 2.5 of this SE for further discussion.

3.5.1.4 Existing Engineering Equivalency Evaluations (EEEEs)

The EEEEs that support compliance with NFPA 805 Chapter 4 were reviewed by the licensee using the methodology contained in NEI 04-02 (Reference 9). The methodology for performing the EEEE review included the following determinations:

• The EEEE is not based solely on quantitative risk evaluations,

• The EEEE is an appropriate use of an engineering equivalency evaluation

• The EEEE is of appropriate quality,

• The standard license condition is met,

• The EEEE is technically adequate,

• The EEEE reflects the plant as-built condition, and

• The basis for acceptability of the EEEE remains valid.

In LAR section 4.2.2, the licensee stated the guidance in RG 1.205, Regulatory Position 2.3.2, (Reference 8) and FAQ 08-0054 (Reference 36) was followed. EEEEs that demonstrate that a fire protection system or feature is "adequate for the hazard" are to be addressed in the LAR as follows:

If not requesting specific approval for "adequate for the hazard" EEEEs, then the EEEE is referenced where required and a brief description of the evaluated condition is provided.

If requesting specific NRC approval for "adequate for the hazard" EEEEs, then the EEEE is referenced where required to demonstrate compliance and is included in LAR Attachment L for NRC review and approval.

The licensee identified and summarized the EEEEs for each fire area in LAR Attachment C, as applicable. The licensee did not request the NRC staff to review and approve any ofthese EEEEs.

Based on the NRC staff's review of the licensee's methodology for review of EEEE's and identification of the applicable EEEEs in LAR Attachment C, the NRC staff concludes that the use of EEEEs meets the requirements of NFPA 805, the guidance of Regulatory Guide 1.205, and FAQ 08-0054, and is acceptable.

3.5.1.5 Variances from Deterministic Requirements For those fire areas where deterministic criteria were not met, VFDRs were identified and evaluated using PB methods. VFDR identification, characterization, and resolutions were identified and summarized in LAR Attachment C for each fire area. Documented variances were all represented as separation issues. The following strategies were used by the licensee in resolving the VFDRs:

- 87

• A FRE determined that applicable risk, DID, and SM criteria were satisfied without further action.

• A FRE determined that applicable risk, DID, and SM criteria were satisfied with a credited RA.

• A FRE determined that applicable risk, DID, and SM criteria were satisfied with a DID-RA.

• A FRE determined that applicable risk, DID, and SM criteria were satisfied with a plant modification(s), as identified in LAR, as supplemented.

For all fire areas where the licensee used the PB approach to meet the NSPC, each VFDR and the associated disposition has been described in LAR Attachment C. Based on the NRC staff review of the VFDRs and associated resolutions as described in LAR Attachment C, as supplemented, the NRC staff concludes that the licensee's identification and resolution of the VFDRs is acceptable.

3.5.1.6 Recovery Actions LAR Attachment G lists the RAs identified in the resolution of VFDRs in LAR Attachment C for each fire area. The RAs identified include both actions considered necessary to meet risk acceptance criteria as well as actions relied upon as DID (see SE Section 3.5.1.7 below).

The NRC staff reviewed LAR Section 4.2.1.3, "Establishing Recovery Actions," and Attachment G, "Recovery Actions Transition," to evaluate whether the licensee meets the associated requirements for the use of RAs per NFPA 805. The details of the NRC staff review for RAs are described in SE Section 3.2.5 "Establishing Recovery Actions." The NRC staff's*

evaluation of the additional risk of RAs credited to meet the risk acceptance guidelines is provided in Section 3.4.4 of this SE.

3.5.1.7 RAs Credited for Defense in Depth The licensee stated in the LAR that RAs required for DID are not credited in the risk determination for the fire area but are credited in the fire risk evaluations. The nuclear safety and radioactive release performance goals, objectives, and criteria of NFPA 805 are met without these actions. These RAs are required for DID and are part of the RljPB FPP which necessitates that these actions would be subject to a plant change evaluation if subsequently modified or removed.

3.5.1.8 Plant Fire Barriers and Separations With the exception of ERFBS, passive fire protection features include the fire barriers used to form fire area boundaries (and barriers separating safe shutdown trains) that were established in accordance with the plant's pre-NFPA 805 deterministic FPP. For the transition to NFPA 805, the licensee decided to retain the previously established fire area boundaries as part of the RIIPB FPP.

- 88 Fire area boundaries are established for those areas described in LAR Attachment C, as rryodified by applicable EEEEs that determine the barriers are adequate for the hazard or otherwise disposition differences in barrier design and performance.from applicable criteria.

The acceptability of fire barriers and separations is also evaluated as part of the NRC staff's review of LAR Attachment A, Table B-1 process and as such are addressed in SE Section 3.1.

3.5.1.9 Electrical Raceway Fire Barrier Systems The licensee stated that the ERFBS used at DAEC met the deterministic requirements of NFPA 805, Chapter 3. Each fire area using ERFBS is identified in LAR Attachment C. There were no VFDRs associated with ERFBS.

3.5.1.10 Conclusion for Section 3.5.1 As documented in LAR Attachment C, for those fire areas that used a deterministic approach in accordance with NFPA 805, Section 4.2.3, the NRC staff concludes that each of the fire areas analyzed using the deterministic approach meet the associated criteria of NFPA 805, Section 4.2.3. This conclusion is based on (1) the licensee's documented compliance with NFPA 805, Section 4.2.3; (2) the licensee's assertion that the success path will be free of fire damage without reliance on RAs;(3) an assessment that the suppression systems in the fire area will have no. impact on the ability to meet the NSPC; and (4) the licensee's appropriate determination of the automatic fire suppression and detection systems required to meet the NSPC.

For those fire areas that used the PB approach in accordance with NFPA 805, S~ction 4.2.4, the NRC staff concludes that each fire area has been properly analyzed, and compliance with the NFPA 805 requirements demonstrated as follows:

Deviations from the pre-NFPA 805 fire protection licensing basis that were transitioned to the NFPA 805 licensing basis were reviewed for applicability, as well as continued validity, and found acceptable.

• VFDRs were evaluated and either found to be acceptable based on an integrated assessment of risk, DID, and SM, or modifications or RAs were identified and actions planned or implemented to address the issue.

RAs used to demonstrate the availability of a success path to achieve the NSPC were evaluated and the additional risk of their use determined, reported, and found to be acceptable. The licensee's analysis appropriately identified the fire protection SSCs required to meet the NSPC, including fire suppression and detection systems.

Fire area boundaries (ceilings, walls, and floors), such as fire barriers, fire barrier penetrations, and through penetration fire stops.

ERFBS credited were documented on a fire area basis, verified to be installed consistent with tested configurations and rated accordingly.

- 89 Accordingly, each fire area utilizing the PB approach was able to achieve and maintain the NSPC, and the associated FREs meet the applicable NFPA 805 requirements for risk, DID, and SMs.

3.5.2 Clarification of Prior NRC Approvals As stated in LAR Attachment T, there are no elements of the current FPP for which NRC clarification is needed.

3.5.3 Fire Protectien during Non-Power Operational Modes (NPO) Modes NFPA 805, Section 1.1, "Scope," states the following:

This standard specifies the minimum fire protection requirements for existing light water nuclear power plants during all phases of plant operation, including shutdown, degraded conditions, and decommissioning.

NFPA 805, Section 1.3.1, "Nuclear Safety Goal," states the following:

The nuclear safety goal is to provide reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition.

The NRC staff reviewed LAR Section 4.3, "Non-Power Operational Modes" and Attachment 0, "NEI 04-02 Table F-1 Non-Power Operational Modes Transition," to evaluate the licensee's treatment of potential fire impacts during NPOs. The licensee followed the guidance used in the process described in NEI 04-02, (Reference 9) as modified by FAQ 07-0040 (Reference 34), for demonstrating that the NSPC are met for higher risk evolutions (HREs) during NPO modes.

3.5.3.1 NPO Strategy and Plant Operating States (POSs)

In LAR Section 4.3 and Attachment 0, the licensee stated that the process used to demonstrate that the I\\ISPC are met during I\\IPO modes is consistent with the guidance contained in FAQ 07-0040 (Reference 34). As described in LAR Attachment 0, the licensee's Nuclear Fleet Guideline outlines use of the "defense-in-depth concept to minimize shutdown risk and

. maximize the availability of critical components and station systems that ensure nuclear safety during shutdown conditions." HREs are "outage activities, plant configurations, or conditions during shutdown where the plant is more susceptible to an event causing the loss of a key safety function." The guideline contains specific actions to address reduced inventory" conditions that consider short time to boil, limited methods for decay heat removal, and low RCS inventory.

As described in the LAR, the licensee identified equipment and cables necessary to support the key safety functions (KSFs) success paths. The operational modes and functional requirements for the systems and components were reviewed. The KSF success path equipment and cables were incorporated in the NPO database model. Following identification of KSF equipment and cables, the licensee performed analysis on a fire area basis to identify areas where redundant equipment and cables credited for a given KSF might fail due to fire damage (Le., pinch-points).

The licensee used a deterministic approach to identify these pinch-points and mitigated these

- 90 pinch-points through the use of RAs and/or fire prevention/protection controls. As stated in Section 4.3.2 of the LAR, fire modeling was not used to eliminate any pinch-points.

I LAR Attachment 0 stated that a NextEra fleet guideline entitled, "Shutdown Risk Management,"

will be used to implement the plant's philosophy of outage risk management for cold shutdown and refueling modes, and when the reactor is defueled. This guidance was developed to provide supplemental site instructions to meet nuclear fleet guidelines for managing shutdown risk. In addition to clarifying definitions and responsibilities, the guideline provides general expectations and specific guidance for maintainiAg each of the KSFs. Prior to entry into Mode 5, and before each planned plant configuration Change, the applicable attachments of the administrative control procedure which implements a shutdown risk management plan, are evaluated for completeness.

In SSA RAI 6a (Reference 41), the NRC staff requested that the licensee identify and describe the changes to outage management procedures, risk management tools, and any other document resulting from incorporation of KSF identified as part of NFPA 805 transition including changes to any administrative procedures such as "Control of Combustibles." In its response to SSA RAI 6a (Reference 13), the licensee stated that they are planning a top-down hierarchical approach to revising NextEra Energy fleet level outage shutdown risk management procedures and the associated plant specific procedures for managing shutdown risk and that these documents will provide departments and organizations that plan outage related work and the plant Risk Assessment Team with shutdown and risk management guidance.

3.5.3.2 NPO Analysis Process The licensee stated that its goal is to ensure that contingency plans are established when the plant is in an HRE and it is possible to lose a KSF due to fire. LAR Section 4.3 discusses these additional controls and measures. However, during low risk periods, normal risk management controls; as well as fire prevention/protection processes and procedures will be used.

In SSA RAI 6b (Part 1) (Reference 41), the NRC staff requested that the licensee provide a list of the additional components for which cable selection was performed and a list of those at power components that have a different functional requirement for NPO. The RAJ also requested that the licensee describe the difference between the at-power safe shutdown function and the NPO function. In its response to SSA RAI 6b (Part 1) (Reference 13), the licensee stated that the NPO Modes Analysis identified systems used for accomplishment of required KSFs and grouped those components making up success paths into function codes.

The licensee further stated that because they were not credited in the at-power analysis, cable selection was performed for 115 electrically-supervised components.

In SSA RAI 6b (Part 2) (Reference 41) the NRC staff requested that the licensee describe the difference between the at-power safe shutdown function and the NPO function. In its response to SSA RAI 6b (Part 2) (Reference 13), the licensee stated that the majority of equipment required to maintain the NPO KSFs is the same as that required to safely shutdown the plant while at power. The licensee further stated that there were 14 electrically supervised safe shutdown components having a different functional requirement during NPO modes; however, since all cables for these components were selected in the current SSA, no additional cable selection was required.

\\

- 91 3.5.3.3 NPO Key Safety Functions and SSCs Used to Achieve Performance LAR Attachment 0 defines the KSFs, the success paths to achieve the KSFs, and the components required for the success paths. In SSA RAI 6c (Reference 41) the NRC staff requested that the licensee provide a list of KSF pinch points by fire area that were identified in the NPO fire area reviews using the method in FAQ 07-0040, including a summary level identification of unavailable paths in each fire area and also describe how these locations will be identified for implementation.

Pinch points refer to a particular location in an area where the damage from a single fire scenario could result in failure of multiples components or trains of a system such that the maximum detriment on that system's performance would be realized from the single fire scenario. Typically, this involves close vertical proximity of cables which support redundant components or trains of a system such that all such cables can be damaged by just one fire scenario.

In its response to SSA RAI 6c (Reference 13), the licensee stated that NPO fire scenarios assumed room/area burnout for the same fire areas evaluated in the at-power analysis and that entire fire areas are identified as KSF pinch points when the NPO fire area review indicates failure of all methods for achieving one or more KSFs. The licensee further stated that in the seven areas listed in the analysis, the assumed NPO fire scenario could cause a loss of all success paths for one or more KSFs. The licensee further stated that each of these areas will be identified through administrative procedures governing fire protection DID features, shutdown risk management, and work control.

In SSA RAI 6d (Reference 41), the NRC staff requested that the licensee provide a description of any actions, including pre-fire staging actions, being credited to minimize the impact of fire induced spurious actuations on power operated valves (POVs) (e.g., air operated valves (AOVs) and motor operated valves (MOVs>> during NPO (e.g., pre-fire rack-out, "pinning" valves, or isolation of air supply). In its response to SSA RAI 6d, (Reference 13), the licensee stated that no particular configuration changes/equipment realignments have been specified to prevent failure of any KSF due to fire during NPO Modes.

In SSA RAI 6e (Reference 41), the NRC staff requested the licensee describe the types of RAs that will be used during normal outage evolutions when certain NPO credited equipment will have to be removed from service. In its response to SSA RAI 6e (Reference13), the licensee stated that additional KSF pinch points introduced by removal of credited equipment from service will be identified through administrative procedures governing fire protection DID features, shutdown risk management, and work control. The licensee further stated that in the unlikely event that such equipment is deliberately removed from service coincident with a planned or emergent HRE, the plant's Risk Assessment Team will consider appropriate contingency measures to reduce fire risk at the additional locations. The licensee further stated that as with pinch points associated with direct fire damage, the proposed options to reduce fire risk will include the same prohibitions and limitations identified in its response to SSA RAI 6a (see discussion above).

- 92 In SSA RAI Sf (Reference 41), the NRC staff requested that the licensee identify those RAs and instrumentation relied upon in NPOs by physical analysis unit and describe how RA feasibility is evaluated. In addition, the NRC staff also requested that the licensee include in the description whether these have been or will be factored into operator procedures supporting these actions.

In its response to SSA RAIS(f) (Reference 13), the licensee stated that no particular operator actions have been specified for restoration of any KSF.

In SSA RAI 8 (Reference 41), the NRC staff noted that in LAR Table B-3, Fire Area CB1, under the column "Required Fire Protection Systems," the licensee stated that the fire brigade response "could be challenging" as the reason for requiring fire detection systems as DID. The NRC staff noted that this phrase is used in numerous entries for numerous fire areas. The NRC staff requested that the licensee clarifies the meaning of "could be challenging" and identify the criteria {or making that determination. In its response to SSA RAI 8 (Reference 12), the licensee stated that each fire zone or physical analysis unit was evaluated by the site to a set of considerations to judge whether the installed fire detection or suppression system( s) are necessary to meet DID in those Fire Areas meeting NFPA 805 Section 4.2.4.2. The licensee further stated that there are no specific criteria for "could be challenging," and that it is a judgment made by the site Fire Marshal, FPP Owner and NSCA engineer. In addition, the licensee stated that the considerations included the following: Class B fire that potentially could spread to multiple elevations, high heat release rates (HHR) fire, limited or constricted avenues of attack, and fire-fighting environments that may be hot, humid or have radiological concerns.

The licensee also stated that the installed detection or suppression system will provide early warning to plant operators to allow time for the fire brigade to arrive on the scene and effectively manage the challenge~ to fire fighting and rapidly extinguish the fire.

. Based on its review of the information provided in the LAR, as supplemented, the NRC staff review concludes that the licensee used methods consistent with the guidance provided in FAQ 07-0040, and RG 1.205 to identify the equipment required to achieve and maintain the fuel in a safe and stable condition during NPO modes. Furthermore, the licensee has a process in place to ensure that fire protection DID measures will be implemented to achieve the KSFs during plant outages. These implementation tasks are reflected in LAR Attachment 0 and S.

3.5.3.4 NPO Pinch Point Resolutions and Program Implementation The licensee identified power operated components needed to support an NPO KSF that all were included in the post-fire safe shutdown equipment list and none required additional circuit analysis. In SSA RAI 7 (Reference 41), the NRC staff requested that the licensee provide a description of what changes are needed to implement the results of the NPO Modes Analysis and where will they be incorporated. In its response to SSA RAI 7 (Reference 13), the licensee stated that the procedures that currently implement shutdown risk and the essential NPO work planning and the implementing processes will be reviewed and modified to implement these changes and requirements.

The licensee further stated that, as described in its response to SSA RAISa, (see discussion above) that a top down hierarchical approach is planned to revising NextEra Energy fleet level outage shutdown risk management procedures and the associated plant specific procedures for managing shutdown risk. In addition, the licensee stated that these documents will provide

- 93 departments and organizations that plan outage related work and the plant Risk Assessment Team with shutdown and risk management guidance.

The analyses performed by the licensee have identified vulnerabilities associated with loss of KSFs during the POSs and associated HREs. In LAR Table S-2, the licensee has identified an action to incorporate the NPO analysis in plant procedures and documentation (Implementation Item 13). LAR Attachment 0 identifies the strategies that will be included in procedures to preclude or mitigate loss of KSFs. By letters dated April 23 and May 23, 2012 (References 12.

and 13), the licensee responded to RAls regarding the NPO discussions in Section 4.3 and of the LAR.

NFPA 805 reql,lires that the NSPC be met during any operational mode or condition, including NPO. As described above, the licensee has performed engineering analyses to demonstrate that it meets this requirement:

Identified the KSFs required to.support the NSPC during NPOs.

Identified the POSs where further analysis is necessary during NPOs.

Identified the SSCs required to meet the KSFs during the POSs analyzed.

Identified the location of these SSCs and their associated cables.

Performed analyses on a fire area basis to identify pinch points were one or more KSFs could be losfas a direct result oUire-induced damage.

Planned/implemented modifications to appropriate procedures in order to employ a fire protection strategy for reducing risk at these pinch points during HREs.

Accordingly, based on the information provided in the LAR, as supplemented, the NRC staff concludes that the licensee has provided reasonable assurance that the NSPC are met during NPO modes and HREs at DAEC.

3.5.4 Conclusion for Section 3.5 The NRC staff reviewed the licensee's RI/PB FPP, as described in the LAR and its supplements, to evaluate the NSCA results. The licensee used a combination of the deterministic approach and the PB approach, in accordance with NFPA 805, Sections 4.2.3 and 4.2.4.

For those fire areas that used a deterministic approach, the NRC staff verified the following:

(I)

The engineering evaluations for exemptions from the existing FPP were evaluated and found to be valid andacceptable for meeting the deterministic requirements of NFPA 805, as allowed by NFPA 805, Section 2.2.7.

Fire suppression effects were evaluated and found to have no adverse impact on the ability to achieve and maintain the NSPC for each fire area.

101 The required automatic fire suppression and automatic fire detection systems were appropriately documented for each fire area.

- 94 Accordingly, the NRC staff concludes that there is reasonable assurance that each fire area utilizing the deterministic approach does so in accordance with NFPA 80S, Section 4.2.3.

For those fire areas that used a PB approach, the NRC staff verified the following:

The engineering evaluations for exemptions from the existing FPP were evaluated and found to be valid and acceptable for meeting the requirements of NFPA 80S, Section 2.2.7.

Fire suppression effects were evaluated and found to have no adverse impact on the ability to achieve and maintain the NSPC for each fire area.

All VFDRs were evaluated using the FRE PB method (in accordance with NFPA 80S, Section 4.2.4.2) to address risk impact, DID, and SM, and were found to be acceptable.

• All RAs necessary to demonstrate the availability of a success path were evaluated with respect to the additional risk presented by their use and found to be acceptable in accordance with NFPA 80S, Section 4.2.4.

The required automatic fire suppression and automatic fire detection systems were appropriately documented for each fire area.

Accordingly, the NRC 13taff concludes that each fire area utilizing the PB approach, in accordance with NFPA 80S, Section 4.2.4, is able to achieve and maintain the NSPC.

Furthermore, there is reasonable assurance that the associated FREs meet the requirements for risk, DID and SM.

The NRC staff's review of the licensee's analysis and outage management process during NPO modes concluded that the licensee provided reasonable assurance that the NSPC will be met during NPO modes and HREs, and that the licensee used methods consistent with the guidance provided in FAQ 07-0040 and RG 1.205. The NRC staff's review also concluded that no RAs are required during NPO modes. The NRC staff concludes that this overall approach for fire protection during NPO modes is acceptable.

3.6 Radioactive Release Performance Criteria NFPA 805 (Reference 7) Chapter 1 defines the radioactive release goals, objectives, and performance criteria that must be met by the fire protection program in the event of a fire at a nuclear power plant.

Radioactive Release Goal.

The radioactive release goal is to provide reasonable assurance that a fire will not result in a radiological release that adversely affects the public, plant personnel, or the environment.

Radioactive Release Objective.

- 95 Either of the following objectives shall be met during all operational modes and plant configurations.

(1) Containment integrity is capable of being maintained (such that fire-fighting products are monitored and released within the plant's normal effluents program), or (2) The source term is capable of being limited (such that any unmonitored releases would not exceed the performance criteria).

Radioactive Release Performance Criteria.

Radiation release to any unrestricted area due to the direct effects of fire suppression activities (but not involving fuel damage) shall be as low as reasonably achievable and shall not exceed applicable 10 CFR Part 20 limits.

In order to assess whether the DAEC FPP to be implemented under NFPA 805 meets the radioactive release goals, objectives, and performance criteria of NFPA 805, the licensee performed an evaluation of the current FPP. The NRC staff performed an audit of the licensee's evaluation of the current FPP to determine whether the DAEC FPP is capable of meeting the NFPA radioactive release goals, objectives, and performance criteria. The results of NRC staff audit are provided below.

The licensee's evaluation used the methodology contained in NEI 04-02 (Reference 9), as endorsed in RG 1.205 (Reference 8), and FAQ 09-0056, "Radioactive Release Transition" (Reference 37). Per FAQ 09-0056, compliance with the radioactive release goals, objectives, and performance criteria can be demonstrated by 1) a review of engineering controls to ensure containment of gaseous and liquid effluents, or 2) otherwise providing a quantitative analysis that demonstrates that the limitations per Technical Specifications are met for instantaneous release of radioactive effluents.

NFPA 805 provides its requirements for engineering analyses to establish the fire protection systems and features necessary to address the nuclear safety and radioactive release goals, objectives and performance criteria. The licensee performed a radioactive release review for all plant operating modes (including pdwer and non-power operations). The licensee's review determined that firefighting activities were defined in the pre-fire plans and fire brigade firefighting instructions the same for any plant operating mode.

For purposes of the fire protection evaluation, the licensee performed an analysis of the potential for effluent releases by subdividing the plant into fire compartments. Each compartment was first screened to determine the potential for generating radioactive effluents during firefighting operations. Compartments where there was no possibility of radioactive materials being present(e.g., those outside of the Radiologically Controlled Area) were eliminated from further review. The licensee provided a detailed summary of its findings on a fire compartment by fire compartment basis as Attachment E, "Radioactive Release Transition,"

of the LAR (Reference 10). The NRC staff's review concluded that the scope of the licensee's review was adequate since each area of the plant was analyzed for its potential for uncontained effluents (see analysis below).

- 96 For compartments where radioactive materials were present (e.g., Reactor Building, Turbine Building, Dffgas Retention Building, Radwaste Building, Low Level Radwaste Processing Facility, and Yard-RCA), the licensee identified the building deSign, engineering controls, administrative controls, fire pre-plans, and fire brigade training materials used to limit potential radioactive release. The engineering controls provide methods of controlling smoke, liquid effluent runoff, and monitoring for radioactive effluents within both gaseous and liquid effluents.

Engineering controls such as structures, installed ventilation, installed drainage and radiation monitoring systems were credited as providing the first level of defense in preventing radioactive release during fire suppression activities.

3.6.1 Gaseous Effluent Controls The Reactor Building ventilation systems, as described in Section 11.5 of the Updated Final Safety Analysis Report (Reference 87), provide a negative pressure throughout potentially contaminated areas of the Reactor Building. The Reactor Building Ventilation is deSigned with radiation monitors that, upon detection of excessive amounts of radioactive material, cause the automatic isolation of secondary containment, stopping all normal ventilation and starting the Standby Gas Treatment Exhaust Systems. The standby ventilation system contains high efficiency particulate filters and charcoal filters to capture radioactive material prior to being released. For monitored release paths, the NRC staff's review determined that the licensee's use of radiation monitoring and automatic ventilation isolation features, combined with effluent filtration, is an adequate method to contain potential effluent releases from monitored release paths to meet instantaneous release rate limits as required by Technical Specifications because the effluent is contained and filtered prior to release. To mitigate unmonitored effluent release paths, the effluent discharges will be contained because the negative atmospheric pressure prevents out-leakage of a potential effluent release.

The Radwaste Building and the Dffgas Retention Building have ventilation systems that discharge to the torus area of the Reactor building, which are then exhausted through the Reactor Building Exhaust System. Gaseous effluents from these buildings will be contained, filtered, and monitored in the same manner as Reactor Building exhaust.

The Turbine Building ventilation system is equipped with radiation monitors and interlocks to stop exhaust in the event radiation levels exceed limits established by the station Technical Specifications. Since the ventilation and radiation monitoring system has the capability to stop the exhaust when a radioactive release is occurring, a release from the turbine building potentially exceeding Technical Specification limits would be stopped and contained.

The Low Level Radwaste Processing Facility ventilation system discharges to an exhaust plenum which is continually monitored for radiation. Any signal from a smoke detector will align dampers to purge the facility through particulate filters, thereby containing a potential release.

For the Yard-RCA, The NRC staff asked Radioactive Release (RR) RAI 1 (Reference 41),

concerning the analysis in this area. In its response (Reference 13), the licensee performed a bounding quantitative analysis of potential radioactive gaseous effluent. Analyses were based on calculational models and assumptions of the worst-case source term (largest amount of radioactive materials) stored in the outside Yard-RCA areas. The NRC staff reviewed the

- 97 models and assumptions and concluded that the assumptions were conservative because they were based on the worst-case (highest) source term and the analytical methods were based on the licensee's Offsite Dose Assessment Manual. The calculational results demonstrated that a fire in the yard-RCA would not result in an offsite dose that exceeded the licensee's Technical Specification instantaneous release rate limits, and thus would be less than the 10 CFR 20 dose limits.

The licensee's review identified other specific plant design features (such as roll-up doors, windows, or drains) that could allow gaseous effluents release. For these potential release paths, the licensee will revise the Area Fire Plans (pre-plans, procedures, and guidelines) to establish administrative controls and guidance for manual actions to be taken by the fire brigade to minimize the offsite releases. The NRC staff's review concludes that effluent releases from other potential release paths will be minimized because administrative controls will be used to further minimize these other potential offsite releases.

3.6.2 Liquid Effluent Controls For liquid effluent control, the licensee performed a review of the control methods for potentially contaminated water runoff from firefighting activities. With the exception of areas identified below, the review identified that floor drains are incorporated into the plant design in most areas of the Radiation Controlled Area and route liquid etnuent to building sumps that are pumped to the Floor Drain Collection Tank. Water in the tank is processed, sampled and either recycled to the Condensate Storage Tank, or discharged in accordance with radwaste procedures to ensure compliance with Technical Specifications for liquid effluent discharges. The NRC staff's review concludes that liquid effluents collected by floor drains are not likely to result in exceeding the 10 CFR 20 limits because the floor drains and sumps provide collection and containment of liquid effluent, and monitoring before discharge to ensure that Technical Specification limits are met.

Other areas in the plant where floor drains do not route to a building sump were identified to drain to the Transformer Deluge Pit in the outside yard. The Deluge Pit, designed as an oil containment system for the Main Power Transformers, and a collection system for oily water from other areas of the plant, also serves as a retention system for these other floor drains. The Deluge Pit is a static containment system located in the yard of the protected area of the plant.

The NRC staff's review concludes that liquid effluent for plant areas where there are no floor drains or sumps will be contained because the liquid effluent collection system (retention pond) is available.

The licensee's review identified other specific plant design features (such as roll-up doors, windows, or drains) that could allow liquid effluent release. For these potential release paths, the licensee will revise the Area Fire Plans (pre-plans, procedures, and guidelines) to establish administrative controls and guidance for manual actions to be taken by the fire brigade to minimize the offsite releases. The NRC staffs review concludes that effluent releases from other potential release paths will be minimized because administrative controls will be used to contain potential liquid releases.

For a liquid effluent release from the Yard-RCA, in its response to RR RAI1 (Reference 13), the licensee performed a bounding quantitative analysis of a potential radioactive liquid release.

- 98 Analyses were based on calculational models and assumptions of the worst-case source term (largest amount of radioactive materials) stored in the outside Yard-RCA areas. The NRC staff reviewed the models and assumptions and concluded that the assumptions were conservative because they were based on the worst-case (highest) source term and reasonable assumptions associated with firefighting activities (e.g., duration of fire, amount of fire water used, and the volume of contaminated fire water washed directly into a storm sewer). The NRC staff concludes that the calculational results demonstrated with reasonable assurance that a fire in the Yard-RCA would not result in exceeding the licensee's Technical Specification instantaneous concentration limits, and thus would be less than the 10 CFR 20 annual dose limits.

The DAEC yard area (areas outside of plant structures and the Yard-RCA) has a storm-water drain system which prevents a direct runoff path into the Cedar River. The storm drains on the southwest corner of the facility, near the Low Level Radwaste Processing Facility route to a retention pond. The retention pond connects to a drainage ditch that connects to the plant drainage canal. A sluice gate is installed at the outlet of the retention pond. This system prevents the inadvertent release of radioactive liquids to the Cedar River from storm drains in this area.

Storm drains for the remainder of the facility (outside the plant structures, Yard-RCA and DAEC yard area), route to one of two outfalls which release directly into the Cedar River. For these remaining areas, the licensee is developing administrative controls such as drain covers and division equipment to minimize the potential for release while performing fire suppression activities. The NRC staff's review concludes that liquid run-off for these outside areas will be prevented because the storm-water drain system collects the direct runoff from the outside yard area and prevents direct runoff into the Cedar River. The NRC staff's review also concludes that the licensee is establishing administrative controls will further minimize the potential for liquid effluent releases for other outside areas.

3.6.3 Pre-Fire Plans The licensee reviewed the existing Pre-Fire Plans and determined that fire brigade actions did not include considerations for the control of radioactivity release. The licensee will amend the Pre-Fire Plans to add text (information) to i.nstruct the Fire Brigade to consider the need for control of radiological release when ventilating smoke or allowing water runoff that could escape from the RCA boundary. The licensee also reviewed Area Fire Plans and drawings to identify areas with exterior doors, and ventilation outlets that may need to be monitored and releases mitigated. The results of these analyses are included in the Attachment E of the LAR. The NRC staff's review concludes that the licensee will reduce the likelihood of effluent releases because Pre-Fire Plans will be amended and used to provide radiological information to the Fire Brigade such that administrative controls can be used when needed to minimize effluent discharges to within 10 CFR 20 dose limits.

3.6.4 Fire Brigade Training Programs The licensee also reviewed the fire brigade training program to determine if the program included objectives to control a radiological release. The current training program does not include the objective of controlling radiological releases to within 10 CFR 20 dose limits, and

- 99 instead relies on the support of the health physics group to address radiological releases. As a result, the licensee identified in Table E-1 of the LAR where several fire pre-plans and training materials need to be revised to provide instructions to identify the potential discharge routes for a radioactive release, provide methods for preventing a potential release and methods of controlling contamination. Fire drills will be used to demonstrate that these objectives can be met. The NRC review concludes that further assurance is given that effluent releases will not result in exceeding the 10 CFR 20 limits because training programs will be revised in order to provide plant staff the knowledge to recognize, minimize and control effluent releases.

The NRC review determined that licensee actions to perform radiation monitoring provide an acceptable method to obtain radiological information for the fire brigade such that, if needed, the brigade can initiate additional administrative controls to limit effluent releases. Since these methods provide containment of effluent releases, the NRC staff concludes that there is reasonable assurance the licensee will meet the NFPA radioactive release objective for gaseous effluents.

3.6.5 Actions to be Completed As indicated in the LAR Table S-2, Item 14, implementation items (such as procedure changes, process updates, Fire Pre-Plans and training) will be completed prior to the implementation of new NFPA 805 FPP program. This will occur within 180 days after NRC approval of the LAR, unless the approval occurs during a scheduled outage window, in which this will occur within 60 days after startup from that scheduled outage.

3.6.6 Conclusion for Section 3.6 The NRC staff have reviewed the licensee's evaluation and concludes that, given the physical design of the plant, the installed engineering controls, and the planned modifications to administrative controls, the licensee has reasonable assurance that it will be able to perform fire suppression activities without exceeding the NFPA 805 radioactive release performance criteria (Le., 10 CFR Part 20 dose limits). Because the licensee's FPP is consistent with the guidance in Regulatory Guide 1.205, and FAQ 09-0056, the NRC staff conclude that there is reasonable assurance that both the instantaneous dose rate limits in the licensee's Technical Specifications and the annual dose limits of 10 CFR 20 will be met:

Thus, based on (1) the information provided in the LAR, as supplemented; (2) the licensee's use of fire pre-plans as' modified; (3) the results of the NRC staff's review of the licensee's evaluation of the identified engineered controls used to manage suppression water and combustion products; and (4) upon completion of the implementation items in Attachment E and Appendix S, "Plant Modifications and Items to be Completed," of the LAR, the development and implementation of revised fire brigade Area Fire Plans (pre-plans, procedures, and guidelines) and training procedures, the NRC staff concludes that the licensee's risk-informed, performance based FPP provides reasonable assurance that radiation releases to any unrestricted area resulting from the direct effects of fire suppression activities are as low as reasonably achievable and are not expected to exceed the NFPA 805 radioactive release performance criteria or the radiological dose limits in 10 CFR Part 20. In conclusion, the NRC staff concludes that the licensee's FPP complies with the requirements specified in NFPA 805, Sections 1.3.2, 1.4.2, and 1.5.2. Accordingly, the NRC staff concludes that this approach is acceptable.

- 100 3.7 NFPA 805 Monitoring Program For this section of the SE, the following requirements from NFPA 805, Section 2.6, are applicable to the NRC staff's review of the DAEC LAR:

NFPA 805 Section 2.6, "Monitoring":

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 FPP 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":

Acceptable levels of availability, reliability, and performance shall be established.

NFPA 805 Section 2.6.2, "Monitoring Availability, Reliability, and Performance":

Methods to monitor availability, reliability, and performance shall be established.

The methods shall consider the plant operating experience and industry operating experience.

NFPA 805 Section 2.6.3, "Corrective Action":

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.

The NRC staff reviewed LAR Section 4.6, "Monitoring Program," that the licensee developed to monitor availability, reliability, and performance of DAEC FPP systems and features after the transition to NFPA 805. The focus of the NRC staff review was on the critical elements related to the monitoring program, including the selection of FPP systems and features to be included in the program, the attributes of those systems and features that will be monitored, and the methods for monitoring those attributes. Implementation of the monitoring program will occur on the same schedule as the NFPA 805 RIIPB FPP implementation, which the NRC staff found acceptable.

The licensee stated that DAEC will develop an NFPA monitoring program consistent with FAQ 10-0059 (Reference 38). Development. of the monitoring program will include a review of existing surveillance, inspection, testing, compensatory measures, and oversight processes for adequacy. The review will examine adequacy of the scope of SSCs within the existing plant programs, performance criteria for availability and reliability of SSCs, and the adequacy of the plant corrective action program. The monitoring program will incorporate phases for scoping, screening using risk criteria, risk target value determination, and monitoring implementation.

The scope of the program will include fire protection systems and features, NSCA equipment,

- 101 -

SSCs relied upon to meet radioactive release criteria, and fire protection programmatic elements.

As described above, NFPA 805, Section 2.6, requires that a mon~oring program be established in order to ensure that the availability and reliability of fire protection systems and features are maintained, as well as to assess the overall effectiveness of the FPP in meeting the performance criteria. Monitoring should ensure that the assumptions in the associated engineering analysis remain valid. Based on the information provided in the LAR, as supplemented, the NRC staff concludes that the licensee's NFPA 805 monitoring program development and implementation process, which is consistent with FAQ 10-0059, provides reasonable assurance that DAEC will implement an effective program for monitoring risk significant fire SSCs because the NFPA 805 monitoring program development and implementation process ensures that the NFPA 805 monitoring program does the following:

Establishes the appropriate performance monitoring groups to be monitored.

Uses an acceptable screening process for determining the SSCs to be included in the performance monitoring groups.

Establishes availability, reliability and performance criteria for the SSCs being monitored.

Requires corrective actions when SSC availability, reliability, and performance criteria targets are exceeded in order bring performance back within the required range.

However, since the final values for availability and reliability, as well as the performance criteria for the SSCs being monitored, have not been established for the monitoring program as of the date of this SE, completion of the DAEC NFPA 805 Monitoring Program is an implementation item, as noted previously (LAR, Attachment S, Plant Modifications and Items to be Completed During Implementation, Table S-2 Implementation Items, Item 2).

Completion of the monitoring program will occur on the same schedule as the implementation of NFPA 805, which the NRC staff concludes is acceptable.

3.7.1 Conclusion for Section 3.7 The NRC staff reviewed the licensee's RI/PB FPP and RAI responses for Section 3.7 of this SE.

The NRC staff concludes that, upon closure of the implementation item in this area, there is reasonable assurance that the licensee's monitoring program meets the requirements specified in Sections 2.6.1,2.6.2 and 2.6.3 of NFPA 805.

3.8 Program Documentation. Configuration Control. and Quality Assurance For this section qf the SE, the requirements from NFPA 805 (Reference 7), Section 2.7, "Program Documentation, Configuration Control and Quality," are applicable to the NRC staff's review of the LAR in regard to the appropriate content, configuration control, and quality of the documentation used to support the DAEC FPP transition to NFPA 805.

- 102 3.8.1 Documentation The NRC staff reviewed LAR Section 4.7.1, "Compliance with Documentation Requirements in Section 2.7.1 of NFPA 805," to evaluate the DAEC FPP'design basis document and supporting documentation.

The DAEC FPP design basis is a compilation of multiple documents (i.e., fire safety analyses, calculations, engineering evaluations, NSCAs, etc.), databases, and drawings which are identified in LAR Figure 4-8, "NFPA 805 Transition - Planned Post-Transition Documentation and Relationships for DAEC." The licensee stated that the analyses conducted to support the NFPA 805 transition were performed in accordance with DAEC processes which meet or exceed the requirements for documentation outlined in NFPA 805, Section 2.7.1.

Specifically, the design analysis and calculation procedures provide the methods and requirements to ensure that design inputs and assumptions are clearly defined, results are easily understood by being clearly and consistently described, and that sufficient detail is.

provided to allow future review of the entire analysis. The process includes provisions for appropriate design and engineering review and approval. In addition, the approved analyses are considered controlled documents, and are accessible via DAEC's document control system.

Being analyses, they are also subject to review and revision consistent with the other plant calculations and analyses, as required by the plant design change process.

The LAR also stated that the documentation associated with the FPP will be maintained for the life of the plant and organized in such a way to facilitate review for accuracy and adequacy by independent reviewers, including the NRC staff.

Based on the LAR description, as supplemented, of the content of the FPP design basis and supporting documentation, and taking into account the licensee's plans to maintain this.

documentation throughout the life of the plant, the NRC staff concludes that the licensee's approach for meeting the requirements of NFPA 805, Sections 2.7.1.1, 2.7.1.2, and 2.7.1.3, regarding adequate development and maintenance of the FPP design basis documentation, is acceptable.

3.8.2 Configuration Control The NRC staff reviewed LAR Section 4.7.2, "Compliance with Configuration Control Requirements in Section 2.7.2 and 2.2.9 of NFPA 805," in order to evaluate the DAEC configuration control process for the new NFPA 805 FPP.

To support the many other technical, engineering and licensing programs at DAEC, the licensee has existing configuration control processes and procedures for establishing, revising, or utilizing program documentation. Accordingly, the licensee is integrating the new FPP design basis and supporting documentation into these existing configuration control processes and procedures. These processes and procedures require that all plant changes be reviewed for potential impact on the various DAEC licensing programs, including the FPP.

The LAR stated that the configuration control process includes; provisions for appropriate design, engineering reviews and approvals, and that approved analyses are considered

- 103 controlled documents available through the DAEC document control system. The LAR also stated that analyses based on the PRA program, which includes the FRE, are issued as formal analyses subject to these same configuration control processes, and are additionally subjected to the PRA peer review process specified in the ASME/ANS PRA standard (Reference 59).

Configuration control of the eXisting FPP during the transition period is maintained by the DAEC change evaluation process, as defined in existing DAEC configuration management and configuration control procedures. DAEC will revise these procedures as necessary for application to the NFPA 805 FPP.

Note that the NRC staff reviewed the licensee's process for updating and maintaining the DAEC FPRA in order to reflect plant changes made after completion of the transition to NFPA 805 is in Section 3.4 of this SE.

Based on the description of the DAEC configuration control process, which indicates that the new FPP design basis and supporting documentation will be controlled documents and that plant changes will be reviewed for impact on the FPP, the NRC staff concludes that there is reasonable assurance that the requirements of NFPA 805 Sections 2.7.2.1 and 2.7.2.2 will be met.

3.8.3 Fire Modeling Quality The NRC staff reviewed LAR Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805," to evaluate the quality of the engineering analyses used to support transition the DAEC FPP to NFPA 805 based on the requirements outlined above. The individual sections of this SE provide the NRC staff's evaluation of the application of the NFPA 805 quality requirements to the licensee's FPP, as appropriate.

3.8.3.1 Review NFPA 805 requires that each analysis, calculation, or evaluation performed be independently reviewed. The licensee stated that its procedures require independent review of analyses, calculations, and evaluations, including those performed in support of compliance with 10 CFR 50.48(c). The LAR also stated that the transition to NFPA 805 was independently reviewed, and that analyses, calculations, and evaluations to be performed post-transition will be independently reviewed, as required by the existing DAEC procedures.

Based on.the licensee's description of the process for performing independent reviews of analyses, calculations, and evaluations, the NRC staff concludes that the licensee's approach for meeting the Quality requirements of NFPA 80S, Section 2.7.3.1, is acceptable.

3.8.3.2 Verification and Validation (V& V)

NFPA 805 requires that each calculational model or numerical method used be verified and validated (V&V'ed) through comparison to test results or other acceptable models. The licensee stated that the calculational models and numerical methods used in support of the transition to NFPA 805 were V&V'ed, and that the calculational models and numerical methods used post transition will be similarly V&V'ed. As an example, the licensee provided extensive information

- 104 related to the V&V of fire models used to support the development of the DAEC FRE. The NRC staff's evaluation of this information is discussed below.

3.8.3.2.1 General NUREG-1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications", Volumes 1-7 (Reference 68), documents the V&V of five selected fire models commonly used to support applications of RII PB fire protection at nuclear power plants. The seven volumes of this NUREG-seri~s report provide technical documentation concerning the predictive capabilities of a specific set of fire dynamics calculation tools and fire phenomenological models that may be used for the analysis of fire hazards in postulated nuclear power plant scenarios. When used within the limitations of the fire models and considering the identified uncertainties, these models may be employed to demonstrate compliance with the requirements of 10 CFR 50.48(c).

Accordingly, for those fire modeling elements performed by the licensee using the V&V applications contained in NUREG-1824 to support the transition to NFPA 805 at DAEC, the NRC staff concludes that the use of these models is acceptable, provided that the intended application is within the appropriate limitations of the model, as identified in NUREG-1824.

In LAR Section 4.5.2, the licensee also identified the use of several empirical correlations that are not addressed in NUREG-1824. The NRC staff reviewed these correlations, as well as the related material provided in the LAR, in order to determine whether the licensee adequately demonstrated alignment with specific portions' of the applicable NUREG-1824 guidance.

Table 3.8-1, "V&V Basis for Fire Modeling Correlations Used at DAEC," in Attachment A and Table 3.8-2, "V&V Basis for Other Fire Models and Related Calculations Used at DAEC," in Attachment B to this SE identify these empirical correlations and algebraic models, respectively, as well as a staff disposition for each.

The NRC staff concluded that the theoretical bases of the models and empirical correlations used in the fire modeling calculations that were not addressed in NUREG-1824 were identified and described in authoritative publications (References 69-80). SE Table 3.8-1 summarizes the additional fire models, and the NRC staff's evaluation of the acceptability of each.

The fire modeling employed by the licensee in the development of the DAEC FRE used empirical correlations that provide bounding solutions for the ZOI; and conservative input parameters, which produced conservative results for the fire modeling analysis. The empiriGal correlations and models were used to develop a generic methodology to determine the ZOI from pre-calculated tables. This methodology is documented in the GFMTs document (Reference 66), as well as what the NRC staff reviewed at the audit. See section 3.4.2.3 for further discussion of the licensee's fire modeling method.

Based on the above, the NRC staff concludes that this approach provides reasonable assurance that the fire modeling used in the development of the fire scenarios for the DAEC FRE is appropriate, and thus acceptable for use in transition to I\\IFPA 805.

- 105 3.8.3.2.2 Discussion of Selected RAI Responses By letters dated January 31, 2012 (Reference 41) and November 8, 2012 (Reference 45), the NRC staff sought additional information (through RAls) concerning the fire modeling conducted to support the fire PRA. By letters dated April 23, 2012 (Reference 12), May 23, 2012 (Reference 13), and January 11, 2013 (Reference 16), the licensee responded to these RAls.

In addition, by letter dated June 25, 2012 (Reference 81), NRC staff transmitted a list of questions that resulted from a second site audit. The purpose of the second audit was to obtain more detailed information for specific fire areas where fire modeling was performed. By letter dated October 15, 2012 (Reference 15), the licensee provided a response to these questions.

The second audit questions were subsequently transmitted to the licensee in the form of RAls by letter dated October 26, 2012 (Reference 82). The following paragraphs describe selected RAI responses related to the acceptability of the fire models used.

(II The NRC staff issued FM RAI 03(a) (Reference 41) to ask the licensee to describe how the empirical correlations in the GFMT document and supplementary material were V&V'ed.

In its response (Reference 12), the licensee explained that the process to develop the GFMTs comprised a development stage, a review stage and an approval stage. A considerable amount of verification was performed during the development stage. The GFMTs tables were developed using Microsoft Excel spreadsheets and Visual Basic macros. These spreadsheets and macros were verified by the preparer, either directly or indirectly through the use of hand or other alternate calculation methods. Additional verification was performed by the reviewer, who followed a standard procedure to review the documents and had access to all calculation materials. The reviewer conducted a detailed review of the implementation of the equation and of the results reported in the GFMTs document. Issues that were identified during the review process were conveyed to the preparer and addressed to the reviewer's satisfaction. The approver conducted a higher level review of the methods, approach and results. Comments from the approver were addressed by the preparer before the final report was released.

The NRC staff concludes that the process to verify the computational tools that were used in the development of the GFMTs document and its supplement, as described in the RAI res'ponse, is acceptable because: (1) the empirical correlation is included in GFMTs for which V& V has been completed and documented in NUREG-1824, and the correlation is applied within the limits of its applicability; (2) the empirical correlation is widely accepted and used by fire protection engineering professionals, is documented in an authoritative publication of the Society of Fire Protection Engineers (SFPE) (e.g., The SFPE Handbook of Fire Protection Engineering). and is applied within the limits of its applicability; or (3) the empirical correlation has been subjected to a peer reView, is published in a widely recognized peer-reviewed journal article or in a conference report (e.g., Fire Safety Journal),

and is applied within the limits of its applicability.

• . The NRC staff issued FM RAI 03(b) (Reference 41) to ask the licensee to; describe the V&V of the empirical equations and correlations identified in the supplement to the GFMTs document, and provide assurance that these equations/correlations were applied within their appropriate range of applicability.

- 106 In its response (Reference 12), the licensee stated that all empirical correlations and algebraic models used in the supplement are also used in the GFMTs document, except for the assumed flame spread rate in cable trays. The latter is based on the generic value for thermoset cables in NUREG/CR-6850 (Reference 62), and is consistent with the test data reported in NUREG/CR-7010 (Reference 83). The response includes a detailed discussion of the validated range of input parameters for the flame height, plume temperature, radiant heat flux and other correlations used in the development of the GFMTs document and its supplement. The response indicates that the correlations were either used within the validated range or in a manner that provides conservative or bounding results.

The NRC staff concluded that the detailed response to RAI 3(b) provides reasonable assurance that the correlations identified in the GFMTs document and its supplement were applied appropriately and therefore are acceptable for transition.

3.8.3.2.3 Post-Transition The licensee also stated that it will revise the appropriate processes and procedures to include NFPA 805 quality requirements for use during the performance of post-transition FPP changes, including those for V& V. Revision of the applicable post-transition processes and procedures to include NFPA 805 requirements for V& V are items in Table S-2 as Implementation Items 15 and 16.

3.8.3.2:4 Conclusion for Section 3.8.3.2 Based on the licensee's description of the DAEC process for V& V of calculational models and numerical methods and their commitment for continued use post-transition, the NRC staff concludes that the licensee's approach to meeting the requirements of NFPA 805 Section 2.7.3.2 is acceptable.

3.8.3.3 Limitations of Use NFPA 805 requires that only acceptable engineering methods and numerical models be used for transition to the extent that the.se methods have been subject to V&V; and that they are applied within the scope, limitations, and assumptions prescribed for that method. The LAR stated that the engineering methods and numerical models used in support of the transition to NFPA 805 were subject to the limitations of use outlined in NFPA 805, Section 2.7.3.3, and that the engineering methods and numerical models used post-transition will be subject to these same limitations of use. As an example, in LAR Sec~ion 4.5.2, "Fire Modeling," the licensee stated that the fire models developed to support the NFPA 805 transition at DAEC fall within their V&V limitations.

3.8.3.3.1 General The NRC staff assessed the acceptability of empirical correlation and fire model in terms of the limits of its use. SE Table 3.8-1 in Attachment A and Table 3.8-2 in Attachment B, summarize the fire models used, how each was applied in the DAEC FRE, the V&V basis for each, and the NRC staff evaluation for each.

- 107 3.8.3.3.2 Discussion of RAls By letters dated January 31, 2012 (Reference 41) and November 8, 2012 (Reference 45), the NRC staff sought additional information (through RAls) concerning the fire modeling conducted to support the fire PRA. By letters dated April 23, 2012 (Reference 12), May 23, 2012 (Reference 13), and January 11, 2013 (Reference 16), the licensee responded to these RAls.

In addition, by letter dated June 25,2012 (Reference 81), NRC staff transmitted a list of questions that resulted from a second site audit. The purpose of the second audit was to obtain more detailed information for specific fire areas where fire modeling was performed. By letter dated October 15, 2012 (Reference 15), the licensee provided a response to these questions.

The second audit questions were subsequently transmitted to the licensee in the form of RAls by letter dated October 26,2012 (Reference 82). The following paragraphs describe selected RAI responses related to the acceptability of the fire models used.

The NRC staff issued FM RAI 04 (Reference 41) to ask the licensee to identify uses of the GFMTs approach outside of the limits of applicability for the method and to explain for those cases how the use of the approach was justified.

In its response (Reference 12), the licensee identified three uses of the GFMTs that may be

outside the limits of applicability of the method: (1) flame height exceeds ceiling height of the compartment; (2) panel dimensions exceed maximum for which lOI tables are applicable; and (3) floor aspect ratio of the compartment where the HGL tables are applied is greater than 5.' For the first case the licensee showed that the horizontal lOI in the GFMTs is conservative. For the second case the licensee determined that the lOI tables in the GFMTs remain conservative for large panels with HRRs up to 783 kW. Since there are no panel fires larger than 717 kW at DAE;C, the pertinent lOI tables in the GFMTs report are conservative. Finally, for the third case only one area was identified at DAEC where the floor area aspect ratio is larger than 5. NUREG-1934 (Reference 88) recommends reducing the volume of the compartment so the aspect ratio is within the range of applicability and then repeat the HGL analysis. The licensee followed this recommendation and the revised HGL analysis was incorporated in the FRE.

Based on its review and the above explanation, the NRC staff concluded that the response to FM RAI 04, provides reasonable assurance that these uses of the GFMTs approach outside its limits of applicability have been adequately addressed and are acceptable for transition.

• During the second audit, the licensee inqicated that it is acceptable to apply the GFMTs approach for panels with dimensions that exceed the upper limit of the analysis, provided the HRR of the panel does not exceed 783 kW. NRC staff issued FM RAI 04.01 (Reference 45) to ask the licensee to explain how this HRR limit would be affected for panels located along a wall or in a corner.

In its response (Reference 15), the licensee noted that GFMTs approach do not explicitly account for wall and corner effects. To account for these effects the "mirror" method is used, i.e., the lOI is determined after increasing the HRR by a factor of two for panels adjacent to a wall, and by a factor of four for panels located near a corner. However, the mirror method also doubles or quadruples the fire area. Hence, the 783 kW limit effectively

- 108 translates to a 1,566 kW limit for panels close to a wall and to 3,132 kW for panels located in a corner.

Based on its review and explanation, the NRC staff concluded that the licensee's adjustment to the limiting HRR for panel fires near a wall or corner is acceptable because the approach is conservative.

3.8.3.3.3 Post-Transition The licensee also stated that it will revise the appropriate processes and procedures to include the NFPA 805 quality requirements for use during the performance of post-transition FPP changes, including those for limitations of use. Revision of the applicable post-transition processes and procedures to include NFPA 805 requirements for limitations of use is an implementation item in Table S-2, numbers 15 and 16.

3.8.3.3.4 Conclusion for Section 3.8.3.3 Based on the licensee's statements that the fire models used to support development of the FRE were used within their limitations, and the description of the DAEC process for placing*

limitations on the use of engineering methods and numerical models, the NRC staff concludes that the licensee's approach to meeting the requirements of NFPA 805 Section 2.7.3.3 is acceptable.

3.8.3.4 Qualification of Users NFPA 805 requires that personnel performing engineering analyses and applying numerical methods (e.g. fire modeling) shall be competent in that field and experienced in the application of these methods as they relate to nuclear power plants, nuclear power plant fire protection, and power plant operations. The licensee's procedures require that cognizant personnel who use and apply engineering analyses and numerical models be competent in the field of application and experienced in the application of the methods, including those personnel performing analyses in support of compliance with 10 CFR 50.48(c).

Specifically, these requirements are being addressed through the implementation of an engineering qualification process at DAEC. The licensee has developed procedures that require that cognizant personnel who use and apply engineering analyses and numerical models be competent in the field of application and experienced in the application of the methods, including those personnel performing analyses in support of compliance with 10 CFR 50.48(c). These requirements are being addressed through the implementation of an engineering qualification process. DAEC has developed qualification or training requirements for personnel performing engineering analyses and numerical methods.

The NRC staff asked RAls pertaining to qualifications of the personnel who supported DAEC FRE fire modeling. Relevant RAls and responses are summarized below:

• The Nf<C staff issued FM RAI 5(a) (Reference 41) to ask the licensee to describe what constitutes the appropriate qualifications for the DAEC staff and consulting engineers who

- 109 use and apply the methods and fire modeling tools included in the engineering analyses and numerical models.

In its response (Reference 12), the licensee stated that the qualification for the leads in the preparation of technical tasks are consistent with and often exceed those articulated in NEI 07-12 for qualification of Peer Reviewers. There are no specific qualifications for the DAEC staff that provide support to the technical leads

• The NRC staff issued FM RAI 5(b) (Reference 41) to ask the licensee to describe the process/procedures for ensuring adequate qualification of the engineers/personnel performing the fire analyses and modeling activities.

In its response (Reference 12), the licensee explained that, since fire modeling required a specialized skill set, acknowledged industry experts were used exclusively for this task.

" The NRC staff issued FM RAI 5(c) (Reference 41) to ask the licensee to explain how the necessary communication and exchange of information between fire modeling analysts and PRA personnel was accomplished.

In its response (Reference 12), the licensee stated that the coordination of technical activities between the fire analysis and risk modeling individuals was facilitated by the availability of a detailed generic fire modeling analysis (Le., GFMT) and a prescriptive set of boundary conditions for which the generic solution would apply. In addition, the fire modeling analyst and the risk modeling individuals were integrated into a single project team which further facilitated and streamlined the communication and exchange of information.

• The NRC staff issued FM RAI 5.01(b) (Reference 45) to obtain additional clarification on procedures in place regarding the above.

In its response (Reference 15), the licensee explained that training and qualification of personnel involved in technical analysis for the DAEC NFPA 805 project is addressed in a "ProjecUQuality Plan" for transition of the existing DAEC FPP to NFPA 805. Per this plan, technical leads are expected to be familiar with the "Project Instruction" document relating to the task or tasks for which they are responsible. Furthermore, with respect to application of the GFMTs approach and the FRE, the technical lead for the project supervised all tasks of the FRE including the integration of the GFMTs into the FRE model. The technical lead was qualified under the DAEC qualification process.

The NRC staff issued FM RAI 5.01(c) (Reference 45) to ask the licensee to identify and describe any procedures used to integrate the process of communication between the PRA and fire modeling groups.

In its response (Reference 15), the licensee stated that no specific procedures or processes were required for communication between the fire modeling group and the PRA group, given the groups were integrated into a single project team. Informal communication was used throughout the project when clarification was required in applying the GFMTs or addressing specific fire modeling concerns outside of the treatments.

- 110 Based on its review and above explanation, the NRC staff concludes that appropriately competent and experienced personnel developed the DAEC FRE, including the supporting fire modeling calculations and including the additional documentation for models and empirical correlations not identified in previous NRC approved V&V documents.

Further, Section 4.7.3 of the tAR, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805 Fire Protection Quality," states that:

... Post-transition, for personnel performing fire modeling or Fire PRA development and evaluation, NextEra Energy Duane Arnold, LLC will develop and maintain qualification requirements for individuals assigned various tasks.

Position Specific Guides will be developed to identify and document required training and mentoring-to ensure individuals are appropriately qualified per the requirements of NFPA 805 Section 2.7.3.4 to perform assigned work...

The post-transition qualification training program will be implemented to include NFPA 805 requirements for Qualification of User in Table S-2 as part of Implementation Item 15.

In addition, based on the licensee's description of the procedures for ensuring personnel who use and apply engineering analyses and numerical methods are competent and experienced, the NRC staff concludes that the licensee's approach for meeting the requirements of NFPA 805, Section 2.7.3.4, is acceptable.

3.8.3.5 Uncertainty Analysis NFPA 805 requires that an uncertainty analysis be performed to provide reasonable assurance that the performance criteria have been met. (Note: 10 CFR 50.48(c)(2)(iv) states that an uncertainty analysis performed in accordance with NFPA 805, Section 2.7.3.5, is not required to.

support calculations used in conjunction with a deterministic approach.) The licensee stated that an uncertainty analysis was performed for the analyses used in support of the transition to NFPA 805, and that an uncertainty analysis will be performed for post-transition analyses.

3.8.3.5.1 General The industry consenSU$ standard for PRA development (Le., the ASME/ANS PRA standard, (Reference 59)) includes requirements to address uncertainty. Accordingly, the licensee addressed uncertainty as a part of the development of the DAEC FRE. The NRC staff's evaluation of the licensee's treatment of these uncertainties is discussed in SE Section 3.4.7.

According to NUREG-1855, Volume 1, "Guidance on the Treatment of Uncertainties Associated with PRAs in RI Decision Making," (Reference 89) there are three types of uncertainty associated with fire modeling calculations:

(1) Parameter Uncertainty: Input parameters are often chosen from statistical distributions or estimated from generic reference data. In either case, the uncertainty of these input parameters affects the uncertainty of the results of the fire modeling analysis.

(2) Model Uncertainty: Idealizations of physical phenomena lead to simplifying assumptions in the formulation of the model equations. In addition, the numerical solution of

- 111 equations that have no analytical solution can lead to inexact results. Model uncertainty is estimated via the processes of V&V. An extensive discussion of quantifying model uncertainty can be found in NUREG-1934, "Nuclear Power Plant Fire Modeling Application Guide (NPP FIRE MAG)." (Reference 88)

(3) Completeness Uncertainty: This refers to the fact that a model is not a complete description of the phenomena it is designed to simulate. Some consider this a form of model uncertainty because most fire models neglect certain physical phenomena that are not considered important for a given application. Completeness uncertainty is addressed by the description of the algorithms found in the model documentation. It is addressed, indirectly by the same process used to address the Model Uncertainty.

3.8.3.5.2 Discussion of Fire Modeling RAls By letters dated January 31,2012 (Reference 41) and November 8,2012 (Reference 45), the NRC staff sought additional information (RAls) concerning the fire modeling conducted to support the fire PRA. By letters dated April 23, 2012 (Reference 12), May 23, 2012 (Reference 13), and January 11, 2013 (Reference 16), the licensee responded to these RAls.

In addition, by letter dated June 25, 2012 (Reference 81), NRC staff transmitted a list of questions that resulted from a second site audit. The purpose of the second audit was to obtain more detailed information for specific fire areas where fire modeling was performed. By letter dated October 15, 2012 (Reference 15), the licensee provided a response to these questions.

The second audit questions were subsequently transmitted to the licensee in the form of RAls by letter dated October 26, 2012 (Reference 82). The following paragraphs describe selected RAI responses related to the acceptability of the fire models used.

The NRC staff issued FM RAI 6( a) (Reference 41) to ask the licensee to explain if and how the results of the sensitivity analyses in the MCR abandonment time study and in the GFMTs report were used in satisfying the requirements of NFPA 805 Section 2.7.3.5.

In its response (Reference 12), the licensee stated that the sensitivity assessments in the GFMTs document and the MCR abandonment time report conclude that conservative inputs were selected, which result in conservative critical separation distances, time to hot gas layer estimates, and time to MCR abandonment estimates. These conservative results are used as the input for identifying fire scenarios.

The NRC staff concluded that the sensitivity analyses in the GFMTs document and the MCR abandonment time report provide reasonable assurance that the ZOI, HGL and MCR abandonment analyses are conservative, and as such, contribute to satisfying the requirements of NFPA 805 Section 2.7.3.5 and acceptable for transition.

3.8.3.5.3 Post-Transition The licensee also stated that it will revise the appropriate processes and procedures to include the NFPA 805 quality requirements for use during the performance of post-transition FPP changes, including those regarding uncertainty analysis. Revision of the applicable post transition processes and procedures to include NFPA 805 requirements regarding uncertainty

- 112 analysis is an implementation item (see Attachment S, Table S-2, Implementation Items, items 15and16).

3.8.3.504 Conclusion for Section 3.8.3.5 Based on the licensee's description of the DAEC process for performing an uncertainty analysis, the NRC staff concludes that the licensee's approach for meeting the requirements of NFPA 805 Section 2.7.3.5 is acceptable.

3.8.3.6 Conclusion for Section 3.8.3 '

Based on the above discussions, the NRC staff concludes that the DAEC RIIPB fire protection quality assurance (QA) program adequately addresses each of the requirements of NFPA 805, Section 2.7.3, which include conducting independent reviews, performing V&V, limiting the application of acceptable methods and models to within prescribed boundaries, ensuring that personnel applying acceptable methods and models are qualified, and performing uncertainty analyses.

3.804 Fire Protection Quality Assurance Program l

GDC 1 of Appendix A to 10 CFR Part 50 requires the following:

Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed.

The licensee established its Fire Protection QA Program in accordance with the guidelines of NUREG-0800, Section 9.5.1 position CA, "Quality Assurance Program," (Reference 57). In addition, the guidance in Appendix C to NEI 04-02 (Reference 9) suggests that the LAR include a description of how the existing fire protection QA program will be transitioned to the new NFPA 805 RI/PB FPP, as discussed below.

The LAR stated that the fire protection QA program is included within and implemented by the DAEC nuclear QA program, although certain aspects of that program are not applicable to the FPP. Further, the LAR stated that no changes to the fire protection QA program were needed to meet the applicable requirements of Section 2.7.3 of NFPA 805. The QA program will be updated in accordance with the requirements of Section 2.7.3 as implementation items are completed per LAR, Table S-2, item 16.

Based on its review and the above explanation, the NRC staff concludes that the licensee's changes to the fire protection QA program are acceptable because they include the expansion of the existing program to include those fire protection systems that were previously not included within the scope of the fire protection QA program that are required by NFPA 805 for transition and post-transition, Table S-2, item 16.

3.8.5 Conclusion for Section 3.8

- 113 The NRC staff reviewed the licensee's RI/PB FPP and RAI responses for Section 3.8 of this SE.

The NRC staff concludes that, upon completion of the implementation items related to the QA program, the licensee's approach for meeting the requirements specified in Section 2.7 of I\\IFPA 805 is acceptable.

4.0 FIRE PROTECTION LICENSE CONDITION The licensee proposed a FPP license condition regarding transition to a RI/PB FPP under NFPA 805, in accordance with 10 CFR 50.48(c)(3)(i). The new license condition adopts the guidelines of the standard fire protection license condition promulgated in RG 1.205, Revision 1, Regulatory Position C.3.1, as issued on December 18,2009 (74 FR 67253). Plant-specific changes were made to the sample license condition; however, the proposed plant-specific FPP license condition is consistent with the standard fire protection license condition, incorporates all of the relevant features of the transition to NFPA 805 at DAEC, and is therefore acceptable.

The following license condition is included in the revised license for the DAEC, and will replace Renewed Operating License No. DPR-49 Condition 2.C(3):

'Fire Protection Program NextEra Energy Duane Arnold, LLC shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated August 5,2011 (and supplements dated October 14,2011, April 23, 2012, May 23, 2012, July 9, 2012, October 15, 2012, January 11, 2013, February 12, 2013, March 6, 2013, May 1, 2013, May 29, 2013, two supplements dated July 2, 2013, and August 5, 2013, and August 28, 2013) and as approved in the safety evaluation report dated September 10, 2013. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the chang~ must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.

- 114 (a) Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins.

The change may be implemented following completion of the plant change evaluation.

(b) Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1 x10-7/year (yr) for CDF and less than 1 x1 0-8/yr for LERF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins.

The change may be implemented following completion of the plant change evaluation.

Other Changes that May Be Made Without Prior NRC Approval

1. Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program.

Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to NFPA 805, Chapter 3 element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard.

The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3 elements are acceptable because the alternative is "adequate for the hazard." Prior NRC review and approval would not be required for alternatives to four specific sections of I'JFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are as follows:

Fire Alarm and Detection Systems (Section 3.8);

• Automatic and Manual Water-Based Fire Suppression Systems (Section 3.9);

Gaseous Fire Suppression Systems (Section 3.10); and Passive Fire Protection Features (Section 3.11).

- 115 This License Condition does not apply to any demonstration of equivalency under Section 1.7 of NFPA 805.

2. Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a

. minimal risk impact. The licensee may use its screening process as approved in the NRC safety evaluation report dated September 10, 2013 to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.

Transition License Conditions (1) Before achieving full compliance with 10 CFR 50.48(c), as specified by (2) and (3) below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (2) above.

(2) The licensee shall implement the modifications to its facility, as described in, Attachment S, Table S-1, "Plant Modifications Committed," of DAEC letter NG-13-0287, dated July 2, 2013, to complete the transition to full compliance with 10 CFR 50.48(c) by December 31,2014. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

(3) The licensee shall implement the items listed in Enclosure 2, Attachment S, Table S-2, "Implementation Items," of DAEC letter NG-13-0287, dated July 2, I

2013, by March 9, 2014.

5.0

SUMMARY

The NRC staff reviewed the licensee's license amendment request, as supplemented, to transition to a RI/PB FPP in accordance with the requirements established by NFPA 805. The NRC staff concludes that the applicant's approach, methods, and data are acceptable to establish, implement and maintain a RI/PB FPP in accordance with 10 CFR 50.48(c).

Implementation of the RIIPB FPP in accordance with 10 CFR 50.48(c) will include the application of.a new fire protection license condition. The new license condition incorporates a list of modifications that must be implemented as well as an established date by which full compliance with 10 CFR 50.48(c) will be achieved. In addition, before the licensee is able to fully implement the transition to a FPP based on NFPA 805 and apply the new fire protection license condition, to its full extent, a number of implementation items must be completed within the timeframe specified.

- 116

6.0 STATE CONSULTATION

In accordance with the Commission's regulations, the State of Iowa official was notified on August 23, 2013 of the proposed issuance of the amendment. The state official had no comments.

7.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on October 4,2011 (76 FR 61393). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22( c)(9).

Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments

8.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations; and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

- 117 Principal Contributors: Paul Lain, NRR Charles Moulton, NRR Jay Robinson, NRR Stephen Dinsmore, NRR Steven Garry, NRR Naeem Iqbal, NRR Bernard Litkett, NRR Daniel O'Neal, NRR Alayna Pearson, NRR Angela Wu, NRR Dennis Andrukat, NRR Karl Bohlander, PNNL Robert Layton, PNNL Eric Schmidt, PNNL Steve Short, PNNL Marc Janssens, CNWRA Robert Fosdick, CNWRA Jason Huczek, CNWRA Patrick Mackin, CNWRA NRR =NRC's Office of Nuclear Reactor Regulation LPNNL = Pacific Northwest National Laboratory CNWRA =Center for Nuclear Waste Regulatory Analysis Date:

Attachments:

A. Table 3.8 V&V Basis for Fire Modeling Correlations Used at DAEC B. Table 3.8 V&V Basis for Fire Model Calculations of Other Models Used at DAEC

- 118

9.0 REFERENCES

1.

10 CFR 50.48, Fire Protection

2.

10 CFR Part 50, Appendix A, Domestic Licensing of Production and Utilization Facilities

3.

10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities

4.

10 CFR 50, Appendix R. Fire Protection Program For Nuclear Power Facilities Operating Prior To January 1, 1979

5.

SECY-98-058, "Development of a Risk-Informed, Performance-Based Regulation for Fire Protection at Nuclear Power Plants," U.S. Nuclear Regulatory Commission, Washington, DC, March 1998 (ADAMS Accession No. ML992910106)

6.

SECY-00-0009, "Rulemaking Plan, Reactor Fire Protection Risk-Informed, Performance Based Rulemaking," U.S. Nuclear Regulatory Commission, Washington, DC, January 2000 (ADAMS Accession No. ML003671923)

7.

NFPA 80S, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants", 2001 Edition, National Fire Protection Association, Quincy, MA

8.

Regulatory Guide 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1, U. S. Nuclear Regulatory Commission, Washington, DC, December 2009 (74 FR 67253; ADAMS Accession No. ML092730314)

9.

NEI 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c)," Revision 2, Nuclear Energy Institute (NEI),

Washington, DC, April 2008 (ADAMS Accession No. ML081130188)

10.

Wells, Peter, NextEra Energy Duane Arnold, LLC, letter to U.S. Nuclear Regulatory Commission, "Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Fire Protection for Light Water Reactor Generating Plants (2001 Edition) (TSCR-128)," dated August 5,2011 (ADAMS Accession No. ML11221A280)

11.

Wells, Peter, NextEra Energy Duane Arnold, LLC, letter to U.S. Nuclear Regulatory Commission, "Clarification of Information Contained in License Amendment Request (TSCR-128): Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition) (TSCR 128)," dated October 14,2011 (ADAMS Accession No. ML112870245)

12.

Wells, Peter,*NextEra Energy Duane Arnold, LLC, letter to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information, License Amendment Request to Adopt National Fire Protection Association Standard 80S, Performance Based Standard For Fire Protection For Light Water Reactor Generating Plants," dated April 23, 2012 (ADAMS Accession No. ML12117A052)

- 119

13.

Wells, Peter, NextEra Energy Duane Arnold, LLC, letter to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information, License Amendment Request to Adopt National Fire Protection Association Standard 805, Performance Based Standard For Fire Protection For Light Water Reactor Generating Plants," dated May 23,2012 (ADAMS Accession No. ML12146A094)

14.

Anderson, Richard L., NextEra Energy Duane Arnold, LLC, letter to U.S. Nuclear Regulatory Commission, "License Amendment Request to Adopt National Fire Protection Association Standard 805, Performance-Based Standard Fqr Fire Protection For Light Water Reactor Generating Plants - Transmittal of CFAST files," dated July 9, 2012 (ADAMS Accession No. ML12200A020)

15.

Anderson, Richard L., NextEra Energy Duane Arnold, LLC, letter to U.S. Nuclear Regulatory Commission, "Response to Second Request for Additional Information, License Amendment Request to Adopt National Fire Protection Association Standard 805, Performance-Based Standard For Fire Protection For Light Water Reactor Generating Plants," dated October 15, 2012 (ADAMS Accession No. ML122910950)

16.

Anderson, Richard L., NextEra Energy Duane Arnold, LLC, letter to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information, License Amendment Request to Adopt National Fire Protection Association Standard 805, Performance-Based Standard For Fire Protection For Light Water Reactor Generating Plants," dated January 11, 2013 (ADAMS Accession No. ML13015A350)

17.

Anderson, Richard L., NextEra Energy Duane Arnold, LLC, letter to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information, License Amendment Request to Adopt National Fire Protection Association Standard 805, Performance-Based Standard For Fire Protection For Light Water Reactor Generatillg Plants," dated February 12, 2013 (ADAMS Accession No. ML13046A031)

18.

Anderson, Richard L., NextEra Energy Duane Arnold, LLC, letter to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information, License Amendment Request to Adopt National Fire Protection Association Standard 805, Performance-Based Standard For Fire Protection For Light Water Reactor Generating Plants," dated March 6, 2013 (ADAMS Accession No. ML13070A065)

19.

Anderson, Richard L., NextEra Energy Duane Arnold, LLC, letter to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information, License Amendment Request to Adopt National Fire Protection Association Standard 80S, Performance-Based Standard For Fire Protection For Light Water Reactor Generating Plants," dated May 1, 2013 (ADAMS Accession No. ML13122A045)

20.

Anderson, Richard L., NextEra Energy Duane Arnold, LLC, letter to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information, License Amendment Request to Adopt National Fire Protection Association Standard 805, Performance-Based Standard For Fire Protection For Light Water Reactor C3enerating Plants," dated May 29,2013 (ADAMS Accession No. ML13150A103)

- 120

21.

Anderson, Richard L., NextEra Energy Duane Arnold, LLC, letter to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information, License Amendment Request to Adopt National Fire Protection Association Standard 805, Performance-Based Standard,For Fire Protection For Light Water Reactor Generating Plants," dated July 2,2013 (ADAMS Accession No. ML13191A035)

22.

Anderson, Richard L., NextEra Energy Duane Arnold, LLC, letter to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information, License Amendment Request to Adopt National Fire Protection Association Standard 805, Performa~nce-Based Standard For Fire Protection For Light Water Reactor Generating Plants," dated July 2,2013 (ADAMS Accession No. ML13189A198)

23.

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24.

NRC Letter from T. A. Ippolito to IELP, (no subject line - Fire Protection Safety Evaluation Report Supplement), February 10, 1981, (ADAMS Accession No. ML021890050)

25.

Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, U. S.

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26.

Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, U. S.

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27.

Regulatory Guide 1.189, "Fire Protection for Operating Nuclear Power Plants," Revision 2, U. S. Nuclear Regulatory Commission, Washington, DC, October 2009 (ADAMS Accession No. ML092580550)

28.

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29.

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30.

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- 121

31.

NRC FAa 07-0030, Establishing Recovery Actions, (ADAMS Accession No. ML110070485)

32.

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33.

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34.

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35.

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36.

NRC FAa 08-0054, Demonstrating Compliance with Chapter 4 of NFPA 805, (ADAMS Accession No. ML110140183)

\\

37.

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38.

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39.

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40.

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41.

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42.

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, 43.

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44.

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45.

NRC E-mail from K. D. Feintuch to DAEC, ME6818 - DAEC Adoption of NFPA-805 Record of RAI clarification conference call conducted on 11/08/2012, November 8,2012, (ADAM S Accession No. ML12318A394)

- 122

46.

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47.

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48.

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49.

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50.

NRC Inspection Report 50-331/03-02 (DRS), May 22,2003, (ADAMS Accession No. ML031430217)

51.

NRC Information Notice 2007-26, "Combustibility of Epoxy Floor Coatings at Commercial Nuclear Power Plants," August 13, 2007

52.

DAEC letter from M. A. Peifer to NRC, "Response to NRC Unresolved Item 50-331103 02-03(DRS): Epoxy Floor Coatings," NG-03-0527, July 25,2003, (ADAMS Accession No. ML032190021)

53.

NRC Inspection Report 50-331/05-09 (DRP), July 12, 2005. (ADAMS Accession No. ML051940049)

54.

American Society of Testing and Materials Standard E84, Standard Test Method for Surface Burning Characteristics of Building Materials, ASTM International, West Conshohocken, Pennsylvania

55.

NEI 00-01, "Guidance for Post Fire Safe Shutdown Circuit Analysis," Revision 2, Nuclear Energy Institute (NEI), Washington, DC, May 2009

56.

NEI 00-01, "Guidance for Post Fire Safe Shutdown Circuit Analysis, Revision 1, Nuclear Energy Institute (NEI). Washington, DC, January 2005

57.

NUREG 0800, Standard Review Plan, Chapter 9.5.1, "Fire Protection Program,"

Revision 3, U. S. Nuclear Regulatory Commission, Washington, DC, July 1981, (ADAMS Accession No. ML052350030)

58.

NRC letter from D. B. Vassallo to IELP, "Safety Evaluation for Appendix R to 10 CFR Part 50, Items III.G.3 and III.L," January 6, 1983, (ADAMS Accession No. ML112240371)

59.

American Society of Mechanical Engineers (ASME) and American Nuclear Society (ANS).ASME/ANS RA-Sa-2009, "Addenda to ASMEIANS RA-S-2008, Standard for

- 123 Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications." ASME, New York, NY. February 2,2009

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U.S. Nuclear Regulatory Commission, "Record of Review, Duane Arnold LAR Attachment U - Table U-1 Internal Events PRA Peer Review - Facts and Observations (F&Os)," dated July 31,2013 (ADAMS Accession No. ML13212A244).

61.

U.S. Nuclear Regulatory Commission, "Record of Review, Duane Arnold LAR Attachment V - Tables V-1 and V-2 Fire PRA Peer Review - Facts and Observations (F&Os)," dated July 31,2013 (ADAMS Accession No. ML13212A257).

62.

NUREG/CR-6850 and EPRI 1011989, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities," Volumes 1 and 2, U. S. Nuclear Regulatory Commission, Washington, DC, September 2005

63.

NRC Letter from Joseph Giitter to Bift Bradley, Nuclear Energy Institute, "Recent Fire PRA Methods Review Panel Decisions And EPRI 1022993, Evaluation Of Peak Heat Release Rates In Electrical Cabinet Fires," June 21,2012 (ADAMS Accession No. ML12171A583)

64.

NRC E-mail from K. D. Feintuch to DAEC, ME6818 - DAEC Adoption of NFPA 805 Request for Additional Information (RAI) Items - Round 2 (More), December 5,2012, (ADAMS Accession No. ML12340A450)

65.

U.S. Nuclear Regulatory Commission, NUREG 1921, "EPRI/NRC-RES Fire Human Reliability Analysis Guidelines," July 2012 (ADAMS Accession No. ML12216A104).

66.

Generic Fire Modeling Treatments, Hughes Associates, Inc., and Kleinsorg Group Risk Services, LCC, Baltimore, MD, January 15, 2008, Revision 0, Non-Publicly Available.

67.

NUREG-1805,"Fire Dynamics Tools (FDTs): Quantitative Fire Hazard Analysis Methods for the U.S. Nuclear Regulatory Commission Fire Protection Inspection Program," U.S. Nuclear Regulatory Commission, Washington, DC, December 2004 (ADAMS Accession No. ML043290075).

68.

NUREG-1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications," Volumes 1 through 7, U. S. Nuclear Regulatory Commission, Washington, DC, May 2007 (ADAMS Accession No. ML062440485).

69.

Shokri, M., and Beyler, C., "Radiation from Large Pool Fires," SFPE Journalof Fire Protection Engineering, Vol. 1, pp. 141-150, 1989.

70.

Mudan, K, "Thermal Radiation Hazards from Hydrocarbon Pool Fires," Progress in Energy and Combustion Science, Vol. 10, pp. 59-80, 1984.

- 124

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Wakamatsu, 1., Hasemi, Y., Kagiya, K., and Kamikawa, D., "Heating Mechanism of Unprotected Steel Beam Installed Beneath Ceiling and Exposed to a Localized Fire:

Verification Using the Real-Scale Experiment and Effects of the Smoke Layer,"

Proceedings of the Seventh International Symposium on Fire Safety Science, International Association for Fire Safety SCience, London, UK, 2003, pp. 1099-1110.

72.

Yokoi, S., "Study on the Prevention of Fire Spread Caused by Hot Upward Current,"

Report Number 34, Building Research Institute, Tokyo, Japan, 1960.

73.

Beyler, C., "Fire Plumes and Ceiling Jets," Fire Safety Journal, Vol. 11, pp. 53-75, 1986.

74.

Gottuk, D. and White, D., "Liquid Fuel Fires," Chapter 2-15, "The SFPE Handbook of Fire Protection Engineering," 3rd Edition, National Fire Protection Association, Quincy, MA, 2002.

75.

Lattimer, B., "Heat Fluxes from Fires to Surfaces," Chapter 2-14, The SFPE Handbook of Fire Protection Engineering, 4th Edition, National Fire Protection Association, Quincy, MA, 2008.

76.

Delichatsios, M., "Flame Heights in Turbulent Wall Fires with Significant Flame Radiation," Combustion Science and Technology, Vol. 39, pp. 195-214, 1984.

77.

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78.

Yuan, L, and Cox, F., "An Experimental Study of Some Line Fires," Fire Safety Journal, Vol. 27, pp. 123-139, 1996.

79.

Lee, B., "Heat Release Rate Characteristics of Some Combustible Fuel Sources in Nuclear Power Plants," NBSIR 85-3196, NBS, Gaithersburg, MD, 1985.

80.

Babrauskas, V., "Estimating Room Flashover Potential," Fire Technology, Vol. 16, pp.94-104, 1980.

81.

NRC E-mail from K. D. Feintuch to DAEC, ME6818 - DAEC Adoption of NFPA 805 Second Audit Visit Questions, June 25, 2012, (ADAMS Accession No. ML12284A021)

82.

NRC E-mail from K.D. Feintuch to DAEC, ME6818 - Additional RAls For DAEC, LAR to Adopt NFPA 805, October 26,2012 (ADAMS Accession No. ML12304A069)

- 125

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84.

NRC Letter from Anthony J. Cappucci to Lee Liu, Iowa Electric Light and Power Company, DAEC Exemption from Appendix R to 10 CFR Part 50 Concerning Separating Redundant Trains By 3 Hour Fire Barriers and Providing Automatic Fire Suppression and Detection Systems, (TAC 55994), October 14, 1987, (ADAMS Accession No. ML021900207).

85.

NRC Letter from Anthony J. Cappucci to Lee Liu, Iowa Electric Light and Power Company, Exemption from Appendic R to 10 CFR Part 50, Concerning Separating Redundant Trains by 3-Hour Fire Barriers and Providing Automatic Fire Suppression and Detection Systems (TAC No. 55994) October 14,1987, (ADAMS Accession No. ML041000504).

86.

NRC Letter from Clyde Y. Shiraki to Lee Liu, Iowa Electric Light and Power Company, Exemption to 10 CFR Part 50, Appendix R, Section III.G.2, (TAC No. 66186), August 16, 1991, (ADAMS Accession No. ML021900627).

87.

DAEC, Updated Final Safety Analysis Report,

88.

NUREG-1934, "Nuclear Power Plant Fire Modeling Analysis Guidelines (NPP FIRE MAG)," U.S. Nuclear Regulatory Commission, Washington, DC, November 2012

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NUREG-1855, Volume 1, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," U. S. Nuclear Regulatory Commission, Washington DC, March 2009

90.

Heskestad, G., "Fire Plumes, Flame Height, and Air Entrainment," Chapter 2-1 of The SFPE Handbook of Fire Protection Engineering, 4th Edition, National Fire Protection Association, Quincy, MA, 2008.

91.

Beyler. C.* "Fire Hazarq Calculations for Large, Open Hydrocarbon Fires." Chapter 3-10 of The SFPE Handbook of Fire Protection Engineering, 4th Edition, National Fire Protection Association, Quincy. MA, 2008.

92.

Peacock, R. Jones, W.* and Reneke. P., "CFAST - Consolidated Model of Fire Growth and Smoke Transport (Version 6) Software Development and Model Evaluation Guide."

NIST Special Publication 1086. National Institute of Standards and Technology, Gaithersburg, MD. 2010.

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Anderson, Richard L.. NextEra Energy Duane Arnold. LLC. letter to U.S. Nuclear Regulatory Commission, "Response to Revised License Condition* for the License

- 126 Amendment Request to Adopt Nationa~1 Fire Protection Association Standard 805, Performance-Based Standard for Fire Protection for Light Water Reactor Generation Plants", dated August 5,2013 (ADAMS Accession No. ML13219A113)

94.

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95.

Anderson, Richard L., NextEra Energy Duane Arnold, LLC, letter to U.S. Nuclear Regulatory Commission, "Response to Revised License Condition for the License Amendment Request to Adopt National Fire Protection Association Standard 805,*

Performance-Based Standard for Fire Protection for Light Water Reactor Generation Plants", dated August 28,2013 (ADAMS Accession No. ML13247A346).

- 127 10.0 ABBREVIATIONS AND ACRONYMS ADAMS AHJ ASC ANS ANSI ASEP ASME ASP ASTM CBDTM CDF CFAST CFR CIV CNWRA CPT CR CRD CST DAEC DHR DID DID-RA EDG EEEE EPRI ERFBS EWS F&O(s)

FAQ FDT FM FPP FPRA FR FRE FSAR GDC GFMT GL Agencywide Documents Access and Management System authority having jurisdiction alternate shutdown capability American Nuclear Society American National Standards Institute accident sequence evaluation program American Society of Mechanical Engineers alternate shutdown panel American Society of Testing and Materials course-based decision tree method core damage frequency consolidated model of fire and smoke transport Code of Federal Regulations containment isolation valves Center for Nuclear Waste Regulatory Analysis control power transformer control room control rode drive condensate storage tank Duane Arnold Energy Center decay heat removal defense-in-depth defense-in-depth recovery action emergency diesel generator existing engineering equivalency evaluations Electric Power Research Institute electrical raceway fire barrier system emergency service water facts and observations frequently asked question fire dynamics simulator fire modeling fire protection program fire probabilistic risk assessment Federal Register fire risk evaluation final safety analysis report general design criteria

.generic fire modeling treatment generic letter

- 128 HGL HPCI HRA HRE HRR HVAC KSF LAR LERF LOOP LPCI MCA MCR MP MSO NEI NFPA NPO NRC NRR NSCA NSPC OMA ORB P&IDs PB PCS POS PRA PSA QA RA RAI RB RCA RCIC RCS RG hot gas layer high-pressure coolant injection human reliability analysis high(er) risk evolution(s) heat release rate heating ventilation and air conditioning key safety function license amendment request large early release frequency loss of offsite power low-pressure coolant injection multi-compartment analysis main control room monitoring program multiple spurious operation Nuclear Energy Institute National Fire Protection Association non-power operation U.S. Nuclear Regulatory Commission (Office of) Nuclear Reactor Regulation nuclear safety capability assessment nuclear safety performance criteria operator manual action offgas retention building Piping and instrumentation drawings performance-based primary control station plant operational state probabilistic risk assessment probabilistic safety assessment quality assurance recovery action request for additional information reactor building radiation controlled area reactor core isolation cooling reactor coolant system regulatory guide

RHR RI RI/PB RIS RB RPV SBLC SE SER SM SR SRV SSA SSC TB TG TR TS UFSAR URI V&V VFDR ZOI

- 129 residual heat removal risk-informed risk-informed, performance-based regulatory issue summary radwaste building reactor pressure vessel standby liquid control safety evaluation safety evaluation report safety margins supporting requirement safety relief valve safe shutdown analysis structures, systems, and components turbine building turbine generator technical report technical specifications updated final safety analysis report unresolved issue verification and validation variance from deterministic requirements zone of influence

I

- A1 Attachment A: Table 3.8-1, V&V Basis for Fire Modeling Correlations Used at DAEC Application at V&V Basis Correlation DAEC Heskestad Development of flame height ZOI tables GFMTs document correlation NUREG-1805 (Reference 67)

NUREG-1824 (Reference 68)

SFPE Handbook (Reference 90)

NRC Staff Evaluation of Acceptability Licensee provided verification of the coding of this correlation in the GFMTs document (Response to FM RAI 3(a),

Reference 12).

• The correlation is validated in NUREG-1824 and an authoritative publication of the Society of Fire Protection Engineers (SFPE) Handbook.

Licensee stated that in most cases, the correlation has been applied within the validated range reported in NUREG-1824.

Licensee provided justification for cases where the correlation was used outside the validated range reported in NUREG 1824 (Response to Fire Modeling RAI 3(b). Reference 12).

Based on its review and explanation, the NRC staff concluded that the use of this correlation in the DAEC application is acceptable..

I

- A2 Attachment A: Table 3.8-1, V&V Basis for Fire Modeling Correlations Used at DAEC r-~~~~~~~~~~

V&V Basis Correlation Application at NRC Staff Evaluation of Acceptability DAEC Heskestad NUREG-1805 Licensee provided verification of the coding of this correlation plume Development of ZOI tables in (Reference 67) in the GFMTs document (Response to FM RAI 3(a),

temperature GFMTs document Reference 12).

NUREG-1824 correlation

• The correlation is validated in NUREG-1824 and an (Reference 68) authoritative publication of the SFPE Handbook.

Licensee stated that in most cases, the correlation has been SFPE Handbook applied within the validated range reported in NUREG-1824.

(Reference 90)

Licensee provided justification for cases where the correlation was used outside the validated range reported in NUREG 1824 (Response to FM RAI 3(b), Reference 12).

Based on its review and explanation, the NRC staff concluded that the use of this correlation in the DAEC application is acceptable.

NUREG-1805

• Licensee provided verification of the coding of this correlation source radiation Modak point Development of (Reference 67)

ZOI tables in in the GFMTs document (Response to FM RAI 3(a),

model GFMTs document Reference 12}.

NUREG-1824 The correlation is validated in NUREG-1824 and an (Reference 68) authoritative publication of the SFPE Handbook Licensee stated that in most cases, the correlation has been SFPE Handbook applied within the validated range reported in NUREG-1824.

(Reference 91)

Licensee provided justification for cases where the correlation was used outside the validated range reported in NUREG 1824 (Response to FM RAI 3(b), Reference 12}.

Based on its review and explanation, the NRC staff concluded that the use of this correlation in the DAEC application is acceptable.

- A3 Attachment A: Table 3.8-1, V&V Basis for Fire Modeling Correlations Used at DAEC Correlation Application at DAEC V&V Basis NRC Staff Evaluation of Acceptability Shokri and Development of Peer-reviewed Licensee provided verification of the coding of this correlation Beyler flame ZOI tables in journal article in the GFMTs document (Response to FM RAI 3(a),

radiation model GFMTs document (Reference 69)

Reference 12}.

The correlation is validated in an authoritative publication.

• Licensee stated that in most cases, the correlation has been applied within the validated range reported in the authoritative publication. Licensee provided justification for cases where the correlation was used outside the reported validated range (Response to FM RAI 3(bJ, Reference 12).

Based on its review and explanation, the NRC staff concluded that the use of this correlation in the DAEC application is acceptable.

Mudan flame Development of Peer-reviewed Licensee provided verification of the coding of this correlation radiation model ZOI tables in GFMTs document journal article (Reference 70) in the GFMTs document (Response to FM RAI 3(a),

Reference 12).

The correlation is validated in an authoritative publication.

Licensee stated that in most cases, the correlation has been applied within the validated range reported in the authoritative publication. Licensee provided justification for cases where" the correlation was used outside the reported validated range (Response to FM RAI 3(b), Reference 12}.

Based on its review and explanation, the NRC staff concluded that the use of this correlation in the DAEC application is acceptable.

- A4 Attachment A: Table 3.8-1, V&V Basis for Fire Modeling Correlations Used at DAEC Correlation Application at DAEC Development of 201 tables in GFMTs document Development of V&V Basis NRC Staff Evaluation of Acceptability Plume heat flux correlation by Wakamatsu et al.

Peer-reviewed conference paper (Reference 71)

National Licensee provided verification of the coding of this correlation in the GFMTs document (Response to FM RAI 3(a),

Reference 12).

• The correlation is validated in an authoritative publication.

Licensee stated that in most cases, the correlation has been applied within the validated range reported in the authoritative publication: Licensee provided justification for cases where the correlation was used outside the reported validated range (Response to FM RAI 3(b), Reference 12).

Based on its review and explanation, the NRC staff concluded that the use of this correlation in the DAEC application is acceptable.

Yokoi plume Licensee provided verification of the coding of this correlation centerline 201 tables in research in the GFMTs document (Response to FM RAI 3(a),

temperature GFMTs document laboratory report Reference 12).

correlation (Reference 72)

Peer-reviewed journal article (Reference 73)

• The correlation is validated in an authoritative publication.

• Licensee stated that in most cases, the correlation has been applied within the validated range reported in the authoritative publication. Licensee provided justification for cases where the correlation was used outside the reported validated range (Response to FM RAI 3(b), Reference 12).

Based on its review and explanation, the NRC staff concluded that the use of this correlation in the DAEC application is acceptable.

- A5 Attachment A: Table 3.8-1, V&V Basis for Fire Modeling Correlations Used at DAEC Correlation Application at V&V Basis NRC Staff Evaluation of Acceptability DAEC Hydrocarbon Development of Licensee provided verification of the coding of this correlation spill fire size SFPE Handbook ZOI tables in (Reference 74) in the GFMTs document (Response to FM RAI 3(a),

correlation GFMTs document Reference 12).

• The correlation is validated in an authoritative publication.

• Licensee stated that in most cases, the correlation has been applied within the validated range reported in the authoritative publication. Licensee provided justification for cases where the correlation was used outside the reported validated range (Response to FM RAI 3(b). Reference 12).

Based on its review and explanation, the NRC staff concluded that the use of this correlation in the DAEC application is acceptable.

.~~~~~~~~

~ ~ ~ ~ ~ ~.

SFPE Handbook

• Licensee provided verification of the coding of this correlation extension Flame Development of (Reference 75)

ZOI tables in in the GFMTs document (Response to FM RAI 3(a),

correlation GFMTs document Reference 12).

• The correlation is validated in an authoritative publication.

• Licensee stated that in most cases, the correlation has been applied within the validated range reported in the authoritative publication. Licensee provided justification for cases where the correlation was used outside the reported validated range (Response to FM RAI 3(b), Reference 12).

Based on its review and explanation, the NRC staff concluded that the use of this correlation in the DAEC application is acceptable.

- A6 Correlation Delichatsios line source flame height model Corner flame height correlation Attachment A: Table 3.8-1,"V&V Basis for Fire Modeling Correlations Used at DAEC Application at V&VBasis NRC Staff Evaluation of Acceptability DAEC Development of Peer-reviE wed Licensee provided verification of the coding of this correlation ZOI tables in journal an cle in the GFMTs document (Response to FM RAI 3(a),

GFMTs document (Reference 76)

Reference 12).

• The correlation is validated in an authoritative publication.

• Licensee stated that in most cases, the correlation has been applied within the validated range reported in the authoritative publicatio[l. Lic!9nsee provided justification for cases where the correlation was used outside the reported validated range (Response to FM RAI 3(b), Reference 12).

Based its review and explanation, the NRC staff concluded that the use of this correlation in the DAEC application is acceptable.

SFPE Handbook Licensee provided verification of the coding of this correlation ZOI tables in Development of

{Referem e 75) in the GFMTs document (Response to FM RAI 3(a),

GFMTs document Reference 12).

• The correlation is validated in an authoritative publication.

Licensee stated that in most cases, the correlation has been applied within the validated range reported in the authoritative publication. Licensee provided justification for cases where the correlation was used outside the reported validated range (Response to FM RAI 3(b), Reference 12).

Based on its review and explanation, the NRC staff concluded that the use of this correlation in the DAEC application is acceptable.

-A7 ""

Attachment A: Table 3.8-1, V&V Basis for Fire Modeling Correlations Used at DAEC Correlation Application at DAEC V&V Basis NRC Staff Evaluation of Acceptability Kawagoe natural vent flow equation Development of 201 tables in GFMTs document National research laboratory report (Reference 77)

Licensee provided verification of the coding of this correlation in the GFMTs document (Response to FM RAI 3(a),

Reference 12).

• The correlation is validated in an authoritative publication.

Licensee stated that in most cases, the correlation has been applied within the validated range reported in the authoritative publication. Licensee provided justification for cases where the correlation was used outside the reported validated range (Response to FM RAI 3(b), Reference 12).

Based on its review and explanation, the NRC staff concluded that the use of this correlation in the DAEC application is acceptable.

Yuan and Cox Development of Peer-reviewed

• Licensee provided verification of the coding of this correlation line fire flame 201 tables in journal article in the GFMTs document (Response to FM RAI ~(a),

height and plume temperature correlations GFMTs document (Reference 78)

Reference 12).

• The correlation is validated in an authoritative publication.

• Licensee stated that in most cases, the correlation has been applied within the validated range reported in the authoritative publication. Licensee provided justification for cases where the correlation was used outside the reported validated range (Response to FM RAI 3(b), Reference 12).

Based on its review and explanation, the NRC staff concluded that the use of this correlation in the DAEC application isacceptable.

- A8 Attachment A: Table 3.8-1, V&V Basis for Fire Modeling Correlations Used at DAEC Correlation v Application at DAEC V&V Basis NRC Staff Evaluation of Acceptability Lee cable fire Development of NBSIR 85-3196 Licensee provided verification of the coding of this correlation model ZOI tables in GFMTs document (Reference 80) in the GFMTs document (Response to FM RAI 3(a),

Reference 12).

• The correlation is validated in an authoritative publication.

• Licensee stated that in most cases, the correlation has been applied within the validated range reported in the authoritative publication. Licensee provided justification for cases where the correlation was used outside the reported validated range (Response to FM RAI 3(b), Reference 12).

Based on its review and explanation, the NRC staff concluded that the use of this correlation in the DAEC application is acceptable.

Babrauskas Development of Peer-reviewed

• Licensee provided verification of the coding of this correlation method to ZOI tables in journal article in the GFMTs document (Response to FM RAI 3(a),

determine GFMTs document (Reference 80)

Reference 12).

ventilation

• The correlation is validated in an authoritative publication.

limited fire size

• Licensee stated that in most cases, the correlation has been applied within the validated range reported in the authoritative publication. Licensee provided justification for cases where the correlation was used outside the reported validated range (Response to FM RAI 3(b), Reference 12).

Based on its review and explanation, the NRC staff concluded that the use of this correlation in the DAEC application is acceptable.

- B1 Attachment B: Table 3.8-2, V&V Basis for Other Fire Models and Related Calculations Used at DAEC Application at V&V Basis Model NRC Staff Evaluation of Acceptability DAEC Development of NUREG-1824,

• The modeling technique is validated in NUREG-1824 (Version 6)

CFAST HGL tables in Volume 5, 2007 (Reference 68) and an authoritative publication of NIST GFMTs document, (Reference 68)

(Reference 92) and MCR licensee stated that in most cases, the correlation has been abandonment NIST Special applied within the validated range reported in NUREG-1824.

times calculations Publication licensee provided justification for cases where the correlation 1086,2008 was used outside the validated range reported in NUREG (Reference 92) 1824 (Response to FM RAI 1, Reference 13).

Based on its review and explanation, the NRC staff concluded that the use of this correlation in the DAEC application is acceptable.

R. Anderson

-2 A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice Sincerely, IRAJ Thomas Wengert for Karl D. Feintuch, Project Manager Plant Licensing Branch 111:-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-331

Enclosures:

1. Amendment No. 286 to DPR-49

2. Safety Evaluation cc w/encls: Distribution via ListServ DISTRIBUTION:

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• concurrence via SE dated 8/112013

• • concurrence via email 8/30/2013 OFFICE NRRlDORULPL3-1/PM NRR/DORULPL3-1/PM NRRlDORULPL4/LA NRRlDRNAPLNBC NRRlDRNAFPB/BC NAME CFaria KFeintuch SRohrer HHamzehee*

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DRoth RCarlson KFeintuch DATE 08/30/13 09/06/13 09/10/13 09/10/13 OFFICIAL RECORD COpy