ML14287A289
| ML14287A289 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 02/11/2015 |
| From: | Shawn Williams Plant Licensing Branch II |
| To: | Gatlin T South Carolina Electric & Gas Co |
| Williams S | |
| References | |
| TAC ME7586 | |
| Download: ML14287A289 (194) | |
Text
UNITED STATES
. NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001.
February 11, 2015 Mr. Thomas D. Gatlin Vice President, Nuclear Operations South Carolina Electric & Gas Company
- Virgil C. Summer Nuclear Station Post Office Box 88, Mail Code 800 Jenkinsville, SC 29065
SUBJECT:
VIRGIL C. SUMMER NUCLEAR STATION, UNIT 1.- ISSUANCE OF AMENDMENT REGARDING TRANSITION TO A RISK-INFORMED, PERFORMANCE-BASED FIRE PROTECTION PROGRAM IN ACCORDANCE WITH 10 CFR 50.48(c) (TAC NO. ME7586)
Dear Mr. Gatlin:
The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 199 to Renewed Facility Operating License. No. NPF-12 for the Virgil C. Summer Nuclear Station, Unit 1 (VCSNS). The amendment consists of changes to the license and Technical Specifications (TSs) in response to your application dated November 15, 2011, as supplemented on January 26. and
_ - October 1 O, 2012;* February 1, April 1, October 14, and November 26, 2013; January 9, February
- 25,* May 2, May 11, August 14, October 9, and December 11, 2014.
The am.endment authorizes the transition of the VCSNS fire protection program to a risk-informed, performance-based program based on National Fire Protection Association (NFPA) 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric*
Generating Plants, 2001 Edition" (NFPA 805), in accordance with Title 10 of the Code of Federal Regulations (1 O 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.
A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's Biweekly Federal Register notice.
Docket No. 50-395
Enclosures:
- 1. Amendment No. 199 to NPF-12
- 2. Safety Evaluation cc w/encls: Distribution via Listserv Sincerely, Shawn Williams, Senior Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SOUTH CAROLINA ELECTRIC & GAS COMPANY SOUTH CAROLINA PUBLIC SERVICE AUTHORITY DOCKET NO. 50-395 VIRGIL C. SUMMER NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 199 Renewed License No. NPF-12
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A The application for amendment to the Virgil C. Summer Nuclear Station, Unit 1, (the facility) Renewed Facility Operating License NPF-12 filed by the South Carolina Electric & Gas Company (the licensee), dated November 15, 2011, as supplemented on January 26 and October 10, 2012; February 1, April 1, October 14, and November 26, 2013; January 9, February 25, May 2, May 11, August 14, October 9, and December 11, 2014, 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 I; 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 as set forth in 10 CFR Chapter I; 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 1 O CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended as indicated in the attachment to this license amendment. Paragraph 2.C.(2) and Paragraph 2.C.(18) of Renewed Facility Operating License No. NPF-12 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 199, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. South Carolina Electric & Gas Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(18)
Fire Protection Program South Carolina Electric and Gas Company 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 11/15/11, (and supplements dated 1/26/12, 10/10/12, 2/1/13, 4/1 /13, 10/14/13, 11/26/13, 1/9/14, 2/25/14, 5/2/14, 5/11/14, 8/14/14, 10/9/14, and 12/11/14) and as approved in the safety evaluation dated February 11, 2015.
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.
- a.
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.
- 1.
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.
j 2.
Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1x10-7 /year (yr) for CDF and less than 1 x1Q-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.
- b.
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 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 3.1 O); 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 dated February 11, 2015, 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.
- c.
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 SCE&G letter RC-14-0196, dated December 11, 2014, by the end of the calendar year 2015. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.
- 3.
The licensee shall implement items listed in Attachment S, Table S-2, "Implementation Items," of SCE&G letter RC-14-0196, dated December 11, 2014, to complete the transition to full compliance with 1 O CFR 50.48(c) by March 31, 2016 as follows:
- a.
Items 3, 6, 7, 8, 10, 13, 14, 17, 19, and 21within180 days of NRC approval.
- b.
Items 1, 2, 4, 11, and 12 by December 31, 2015.
- c.
Items 5, 15, 16, 18, 20, 22, and 23 by March 31, 2016.
- 3.
This amendment is effective as of its date of issuance and shall be implemented per the December 11, 2014, supplement, Attachment S, Table S-2 "Implementation Items",
requiring full implementation by March 31, 2016.
Attachment:
Changes to Renewed Facility Operating License No. NPF-12 FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: February 11, 2015
ATTACHMENT TO LICENSE AMENDMENT NO. 199 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-12 DOCKET NO. 50-395 Replace the following pages of the Renewed Facility Operating License and Appendix "A" Technical Specification pages with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Pages License NPF-12, Page 3 NPF-12, Page 7 NPF, Page 8 TS 6-11 Insert Pages License NPF-12, Page 3 NPF-12, Page 7 NPF-12, Page 7a NPF-12, Page 7b NPF-12, Page 8 TS 6-11 (3)
SCE&G, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage amounts required for reactor operation, as described in the Final Safety Analysis Report, as amended through Amendment No. 33; (4)
SCE&G, pursuant to the Act and 10 CFR Part 30, 40 and 70 to receive, possess and use at any time byproduct, source and special nuclear material as seated neutron sources for reactor startup, sealed neutron sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)
SCE&G, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess and use in amounts as required any byproduct source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus of components; and (6)
SCE&G, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess,, but not separate, such byproduct and special nuclear materials as my be produced by the operation of the facility.
C.
This renewed license shall be deemed to contain, and is subject to, the conditions specified in the Commission's regulations set forth in 1 O CFR Chapter I and 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 SCE&G is authorized to perathe the facility at reactor core power levels not in excess of 2900 megawatts thermal in accordance with the conditions specified herein and in Attachment 1 to this renewed license.
The preoccupation tests, startup tests and other items identified in to this renewed license shall be completed as specified. is hereby incorporated into this renewed license.
(2)
Technical Specifications and Environmental Protection Plant The Technical Specifications contained in Appendix A, as revised through Amendment No.199 and the Environmental Protection Plan contained in Appendix 8, are hereoy incorporated in the renewed license. South Carolina Electric & Gas Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
Renewed Facility Operating License No. NPF-12 Amendment No.199
- b.
In the event that one-third thickness semi-circular reference flaws cannot be detected and discriminated from inherent anomalies, the entire volume of the weld shall be examined during the inservice inspection.
- c.
The reporting of the inservice inspection examination results shall be documented in a manner to define qualitatively whether, the weldment and the heat affected zone and adjacent base metal on both sides of the weld were examined by ultrasonic angle beam techniques.
(9)
Design Description - Control (Section 4.3.2. SER)
SCE&G is prohibited from using part-length rods du-ring power operation.
(13)
Deleted (14)
Deleted (15)
Deleted (16)
Cable Tray Separation (Section 8.3.3, SSER 4)
Prior to startup after the first refueling outage, SCE&G shall implement the modifications to the cable trays discussed in Section 8.3.3 of Supplement No. 4 to the Safety Evaluation Report or demonstrate to the NRC staff that faults induced in non-class 1 E cable trays will not result in fajlure of cable in the adjacent Class 1 E cable trays.
(17)
Alternate Shutdown System Section 9.5.1, SSER 4)
Prior to startup after the first refueling outage, SCE&G shall install a source range neutron flux monitor independent of the control complex as part of the alternate shutdown system.
(18)
Fire Protection Program South Carolina Electric & Gas Company shall implement and maintain in effect all provisions of the approved fire protection program that comply with 1 O CFR 50.48(a) and 1 O CFR 50.48(c), as specified in the licensee amendment request dated 11 /15/11 (and supplements dated 1/26/12, 10/10/12, 2/1 /13, 4/1 /13, 10/14/13, 11/26/13, 1 /9/14, 2/25/14, 5/2/14, 5/11 /14, 8/14/14, 10/9/14, and 12/11 /14) and as approved in the safety evaluation report dated 02/11 /15.
__ 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 1 O CFR 50.48(a) and 1 O 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 Facility Operating License No. NPF-12
-?a-
- a.
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.
- 1.
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.
- 2.
Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1x10-7/year (yr) for CDFand less than 1x10~
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.
- b.
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 approve 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.
Renewed Facility Operating License No. NPF-12
-7b-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 tor the hazard. A qualified fire protection engineer shall approve 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 3.1 O); and, Passive Fire Protection Features (Section 3.11 ).
This License Condition does not apply to any demonstrati.on 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 aooroved in the NRC safety evaluation dated February 11, 2015.
The licensee shall ensure that fire' protection aefense-in-depth and safety margins are maintained when changes are made to the fire protection program.
- c.
Transition License Conditions
- 1.
Before achieving full compliance with 1 O CFR 50.48(c), as specified by c.2 and c.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 descri.bed in b.2 above.
- 2.
The licensee shall implement the modifications to its facility,.
as described in Attachment S, Table S-1, "Plant Modifications Committed," of SCE&G letter RC-14-0196, dated December 11, 2014, by the end of the calendar year 2015. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.
Renewed Facility Operating License No. NPF-12
- 3.
The licensee shall implement items listed in Attachment S, Table S-2, "Implementation Items," of SCE&G letter RC-14-0196, dated December 11, 2014, to complete the transition to full compliance with 1 O CFR 50.48(c) by March 31, 2016 as follows:
- a.
Items 3, 6, 7, 8, 10, 13, 14, 17, 19, and 21within180 days of NRC approval.
- b.
Items 1, 2, 4, 11, and 12 by December 31, 2015.
- c.
Items 5, 15, 16, 18; 20, 22, and 23 by March 31, 2016.
(19)
Instrument and Control Vibration Tests for Emergency Diesel Engine Auxiliary Support Systems (Section 9.5.4, SER)
Prior to startup after the first refueling outage, SCE&G shall either provide test results and results of analyses to the NRC staff for review and approval which validate that the skid-mounted control panels and mounted equipment have been developed, tested, and qualified for operation under severe vibrational stresses encountered during diesel engine operation, or SCE&G shall floor mount the control panels presently furnished with the diesel generators separate from the skid on a vibration-free floor area.
(20)
Solid Radioactive Waste Treatment System (Section 11.2.3, SSER 4)
SCE&G shall not ship "wet" solid wastes from the facility until the NRC staff has reviewed and approved the process control program for the cement solidification system.
(21)
Process and Effluent Radiological Monitoring and Sampling Systems (Section 11.3, SSER 4)
Prior to startup after the first refueling outage, SCE&G shall install and calibrate the condensate demineralizer backwash effluent monitor RM-L 11.
(22)
Core Reactivity Insertion Events (Section 15.2.4, SSER 4)
For operations above 90% of full power, SCE&G shall control the reactor manually or the rods shall be out greater than 215 steps until written approval is received from the NRC staff authorizing removal of this restriction.
(23)
NUREG-0737 Conditions (Section 22)
SCE&G shall complete the following conditions to the satisfaction of the NRC staff. Each item references the related subpart of Section 22 of the SER and/or its supplements. *
- a.
Procedures for Transients and Accidents (l.C.1. SSER 4)
Prior to startup after the first refueling outage, SCE&G shall implement emergency operating procedures based on guidelines approved by the NRC staff.
Renewed Facility Operating License No. NPF-12
ADMINISTRATIVE CONTROLS
- d.
Critical operation of the unit shall not be resumed until authorized by the Commission.
6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:
- a.
The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.
- b.
Refueling operations.
- c.
Surveillance and test activities of safety-related equipment.
- d.
Security Plan.
- e.
- f.
- g.
OFFSITE DOSE CALCULATION MANUAL.
- h.
Effluent and environmental monitoring program using the guidance in Regulatory Guide 4.15, Revision 1, February 1979.
6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be reviewed prior to implementation as set forth in 6.5 above.
6.8.3 NOT USED.
6.8.4 The following programs shall be established, implemented and maintained:
- a.
Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the chemical and volume control, letdown, safety injection, residual heat removal,
. nuclear sampling, liquid radwaste handling, gas radwaste handling and reactor building spray system. The program shall include the following:
- 1)
- 2)
Preventive maintenance and periodic visual inspection requirements, and Integrated leak test requirements for each system at refueling cycle intervals or less.
- b.
In-Plant Radiation Monitoring
- 1)
Training of personnel,
- 2)
Procedures for monitoring, and
- 3)
Provisions for maintenance of sampling and analysis equipment.
SUMMER - UNIT 1 6-11 Amendment No.199
ENCLOSURE 1 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. 199 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-12 SOUTH CAROLINA ELECTRIC & GAS COMPANY VIRGIL C. SUMMER NUCLEAR STATION. UNIT 1 DOCKET NO. 50-395
TABLE OF CONTENTS SAFETY EVALUATION Contents
1.0 INTRODUCTION
......................................................................................................... 1.1 Background................................................................................................................ 1.2 Requested Licensing Action.......................................................................................
2.0 REGULATORY EVALUATION
................................................................. :................. 2.1 Applicable Regulations............................................................................................... 2.2 Applicable Staff Guidance.......................................................................................... 2.3 NFPA 805 Frequently Asked Questions********************************'********************************** 2.4 Orders, License Conditions and Technical Specifications......................................... 2.4.1 Orders.. :............................................................'....................................................... 2.4.2 License Conditions................................................................................................... 2.4.3 Technical Specifications............................................................................... 2.4.4 Updated Final Safety Analysis Report...................................................................... 2.5 Rescission of Exemptions........................................................................................ 2.6 Self Approval Process for Fire Protection Program Changes (Post-Transition)........ 2.6.1 Post-Implementation Plant Change Evaluation Process........................................... 2.6.2 Requirements for the Self Approval Process Regarding Plant Changes................... 2.7 Modifications and Implementation Items................................................................... 2.7.1 Modifications............................................................................................................ 2.7.2 Implementation Items............................................................................................... 2.7.3 Schedule............................................................ :.....................................................
3.0 TECHNICAL EVALUATION
..................................................................................... 3.1 NFPA 805 Fundamental Fire Protection Program (FPP) and Design Elements........ 3.1.1 Compliance with NFPA 805 Chapter 3 Requirements.............................................. 3.1.1.1 Compliance Strategy -- Complies.................................................................. 3.1.1.2 Compliance Strategy -- Complies by Alternative............................................ 3.1.1.3 Compliance Strategy -- Complies with Use of FPEEEs.................................. 3.1.1.4 Compliance Strategy -- Complies via Previous NRG Approval....................... 3.1.1.5 Compliance Strategy - Submit for NRG Approval.......................................... ~ 34 -
3.1.1.6 Compliance Strategy - No Review Required................................................. 3.1.1. 7 Compliance Strategy -- Multiple Strategies.................................................... 3.1.1.8 Chapter 3 Sections Not Reviewed................................................................. 3.1.1.9 Compliance with Chapter 3 Requirements Conclusion................................... 3.1.2 Identification of Power Block.................................................................................... 3.1.3 Closure of Generic Letter 2006-03, "Potentially Nonconforming Hemyc' '
and MT' Fire Barrier Configurations" Issues........................................................... 3.1.4 Performance-Based Methods for NFPA 805 Chapter 3 Elements............................ 3.1.4.1 Use of Non-treated Wood in Limited Quantities............................................. 3.1.4.2 Limited Unqualified Wiring Above Suspended Ceilings.................................. 3.1.4.3 Electrical Cable Construction with Non-compliant Flame Propagation Tests.. 3.1.4.4 Bulk Gas Storage Tanks................................................................................
3.1.4.5 Fire Brigade Notification................................................................................. 3.1.4.6 Pre-Fire Plans................................................................................................ 3.1.4.7 Records............................................................................................................ 3.1.4.8 Yard Fire Hydrant Layout............................................................................... 3.1.4.9 Hose Stations - Pressure Reducers................................................. *............. 3.1.4.10 Seismic Analyzed Hose Stations........................................................... 3.1.4.11 Fire Detection Code of Record.............................................................. 3.1.4.12 Reactor Coolant Pumps........................................................................ 3.1.4.13
. Procedures................................................................................ :........... 3.2 Nuclear Safety Capability Assessment Methods................................,..................... 3.2.1 Compliance with NFPA 805 Nuclear Safety Capability Assessment Methods.......... 3.2.1.1 Attribute Alignment -- Aligns........................................................................... 3.2.1.2 NFPA 805 Nuclear Safety Capability Assessment Methods Conclusion........ 3.2.2 Maintaining Fuel in a Safe and Stable Condition...................................................... 3.2.3 Applicability of Feed and Bleed................................................................................ 3.2.4 Assessment of Multiple Spurious Operations........................................................... 3.2.5 Establishing Recovery Actions................................................................................. 3.2.6 Plant-Specific Treatments or Technologies (Incipient Fire Detection Systems)........ 3.2.7 Conclusion for Section 3.2........................................................................................ 3.3 Fire Modeling Performance-Based Approach........................................................... 3.3.1 Overview of the Fire Modeling Performance-Based Approach.................................. 3.3.2 Defense-in-Depth..................................................................................................... 3.3.3 Safety Margins......................................................................................................... 3.3.4 Fire Models Used in the Analysis............................................................................. 3.3.5 RAls Pertaining to the Performance-Based Fire Modeling Approach........................ 3.3.6 Conclusion for Section 3.3............ :........................................................................... 3.4 Fire Risk Evaluations................................................................................................ 3.4.1 Maintaining Defense-in-Depth and Safety Margins................................................... 3.4.1.1 Defense-in-Depth (DID)................................................................................. 3.4.1.2 Safety Margins............................................................................................... 3.4.2 Quality of the Fire Probabilistic Risk Assessment.................................... *.*............... - 82 ::-
3.4.2.1 Internal Events PRA Model............................................................................ - 82 *_
3.4.2.2 Fire PRA Model............................................................................................. 3.4.2.3 Fire Modeling in Support of the Development of the Fire Risk Evaluations (FREs)........................................................................................................... 3.4.2.3.1 Overview of Fire Models Used to Support the Fire Risk Evaluations.... 3.4.2.3.2 RAls Pertaining to Fire Modeling in Support of the VCSNS FREs........ 3.4.2.3.3 Conclusion for Section 3.4.2.3............................................................ - 102 -
3.4.2.4 Conclusions Regarding Fire PRA Quality..................................................... - 102 -
3.4.3 Fire Risk Evaluations.............................................................................................. - 102 -
3.4.4 Additional Risk Presented by Recovery Actions..................................................... - 104 -
3.4.5 Risk-Informed or Performance-Based Alternatives to Compliance with NFPA 805. - 105 -
3.4.6 Cumulative Risk and Combined Changes.............................................................. - 105 -
3.4. 7 Uncertainty and Sensitivity Analyses...................................................................... - 107 -
3.4.8 Conclusion for Section 3.4...................................................................................... - 108 -
3.5 Nuclear Safety Capability Assessment Results...................................................... - 109 -
3.5.1 Nuclear Safety Capability Assessment Results by Fire Area.................................. - 109 -
3.5.1.1 Fire Detection & Suppression Systems Required to Meet the Nuclear Safety Performance Criteria.................................................................................... - 114 -
3.5.1.2 Evaluation of Fire Suppression Effects on Nuclear Safety Performance Criteria......................................................................................................... - 115 -
3.5.1.3 Licensing Actions......................................................................................... - 115 -
3.5.1.4 Existing Fire Protection Engineering Equivalency Evaluations (FPEEEs).... - 124 -
3.5.1.5 Variances from Deterministic Requirements (VFDRs).................................. - 125 -
3.5.1.6 Recovery Actions......................................................................................... - 126-3.5.1.7 Recovery Actions Credited for Defense in Depth (RA-DID).......................... - 126-3.5.1.8 Plant Fire Barriers and Separations............................................................. - 126 -
3.5.1.9 Electrical Raceway Fire Barrier Systems (ERFBS)...................................... - 127 -
3.5.1.10 Issue Resolution................................................................................. -127 -
3.5.1.11 Conclusion for Section 3.5.1............................................................... - 128 -
3.5.2 Clarification of Prior NRC Approvals....................................................................... - 129 -
3.5.3 Fire Protection During Non-Power Operational Modes........................................... - 130 -
3.5.3.1 NPO Strategy and Plant Operational States (POSs).................................... - 130 -
3.5.3.2 NPO Analysis Process...........,..................................................................... - 131 -
3.5,3.3 NPO Pinch Point Resolutions and Program Implementation........................ - 131 -
3.5.3.4 Conclusion for NPO..................................................................................... - 132 -
3.5.4 Conclusion for Section 3.5...................................................................................... - 133 -
3.6 Radioactive Release Performance Criteria............................................................. - 134 -
3.6.1 Method of Review................................................................................................... - 135 -
3.6.2 Scope of Review.................................................................................................... - 135 -
3.6.3 Identification of Plant Areas Containing Radioactive Materials...................... :........ - 136 -
3.6.4 Fire Pre-Plans........................................................................................................ - 136 -
3.6.5 Gaseous Effluent Controls...................................................................................... - 137 -
3.6.6 Liquid Effluent Controls.......................................................................................... - 138 -
3.6.7 Fire Brigade Training Materials............................................................................... - 139 -
3.6.8 Conclusion........................ -............................ :........................................................ - 139 -
3.7 NFPA 805 Monitoring Program............................................................................... - 140 -
3.7.1 Conclusion for Section 3.7...................................................................................... - 141 -
3.8 Program Documentation, Configuration Control, and Quality Assurance................ - 142 -
3.8.1 Documentation....................................................................................................... - 143 -
3.8.2 Configuration Control............................................................................................. - 144 -
3.8.3 Quality.................................................................................................................... - 145 -
3.8.3.1 Review......................................................................................................... - 145 -
3.8.3.2 Verific~tion and Validation (V&V)................................................................. - 145 -
3.8.3.2.1 General.............................................................................................. - 145 -
3.8.3.2.2 Discussion of RAI Responses............................................................ - 146 -
3.8.3.2.3 Post-Transition................................................................................... - 148 -
3.8.3.2.4 Conclusion for Section 3.8.3.2............................................................ - 148 -
3.8.3.3 Limitations of Use........................................................................................ - 148 -
3.8.3.3.1 General.............................................................................................. - 149 -
3.8.3.3.2 Discussion of RAls............................................................................. - 149 -
3.8.3.3.3 Post-Transition................................................................................... - 149 -
3.8.3.3.4 Conclusion for Section 3.8.3.3.......................................,.................... - 149 -
3.8.3.4 Qualification of Users................................................................................... - 150 -
3.8.3.5 Uncertainty Analysis*.................................................................................... - 150 -
3.8.3.5.1 General..........................................'.................................................... - 151 -
3.8.3.5.2 Discussion of RAls........................ :.................................................... - 151 -
3.8.3.5.3 Post-Transition................................................................................... - 152 -
3.8.3.5.4 Conclusion for Section 3.8.3.5............................................................ - 152 -
3.8.3.6 Conclusion for Section 3.8.3........................................................................ - 152 -
3.8.4 Fire Protection Quality Assurance Program............................................................ - 152 -
3.8.5 Conclusion for Section 3.8................ :..................................................................... - 153 -
4.0 FIRE PROTECTION LICENSE CONDITION........................................................... -153 -
5.0
SUMMARY
............................................................................................................. - 156 -
6.0 STATE CONSULTATION
....................................................................................... -156 -
7.0 ENVIRONMENTAL CONSIDERATION
.................................................................. -157 -
8.0 CONCLUSION
.......................................................... :............................................. - 157 -
9.0 REFERENCES
........................................................................................................ - 157 -
Attachment A: Table 3.8-1, V&V Basis for Fire Modeling Correlations Used at VCSNS....................................................................................................................... Attachment B: Table 3.8-2, V&V Basis for Other Models Used at VCSNS....................... Attachment C: Abbreviations and Acronyms
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 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. 199 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-12 SOUTH CAROLINA ELECTRIC & GAS COMPANY VIRGIL C. SUMMER NUCLEAR STATION, UNIT 1 DOCKET NO. 50-395
1.0 INTRODUCTION
1.1 Background
The U.S. Nuclear Regulatory Commission (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" and Appendix R 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 and states that the fire protection plan must describe the overall fire protection program (FPP); 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, performance-based (RI/PB), consensus standard for fire protection. In 2001, the NFPA Standards Council issued NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants" (Reference 1),
which describes a 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 (RG) 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1, (Reference 2), states:
On March 26, 1998, the NRC staff sent to the Commission SECY-98-058, "Development of a Risk-Informed, Performance-Based Regulation for Fire Protection at Nuclear Power Plants" [Reference 3], 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 4], 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," as an American National Standard for performance-based fire protection for light-water nuclear power plants.
A licensee 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 licensee then must establish plant fire protection requirements using the methodology in Chapter 2 of NFPA 805 such that the minimum fire protection program elements and design criteria contained in Chapter 3 of NFPA 805 are satisfied. Next, the licensee 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, the licensee will use engineering evaluations, probabilistic S?tfety 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, in part, 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 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 fire protection program 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, in part, that:
In parallel with the Commission's efforts to issue a rule incorporating the risk-informed, performance-based fire protection provisions of NFPA 805, NEI [the Nuclear Energy Institute] published implementing guidance for the *specific provisions of 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)," Revision 2 (Reference 5)].
RG 1.205 provides the NRC staffs position on NEI 04-02, Revision 2 (Reference 5), and offers additional *information and guidance to supplement the NEI document and assist licensees in meeting the NRC's regulations in 10 CFR 50.48(c) related to adopting a risk-informed, performance-based fire protection program. RG 1.205 endorses the guidance of NEI 04-02, Rev. 2, subject to certain exceptions, as providing methods acceptable to the staff for adopting a fire protection program consistent with the 2001 edition of NFPA 805 and 10 CFR 50.48(c).
Accordingly, South Carolina Electric & Gas Company (SCE&G), requested a license amendment to allow the licensee to maintain the Virgil C. Summer Nuclear Station (VCSNS) Unit 1 fire protection program in accordance with 10 CFR 50.48(c).
1.2 Requested Licensing Action By letter dated November 15, 2011 (Reference 6), as supplemented by letters dated January 26, 2012 (Reference 7), October 10, 2012 (Reference 8), February 1, 2013 (Reference 9), April 1, 2013 (Reference 10), October 14, 2013 (Reference 11 ), November 26, 2013 (Reference 12),
January 9, 2014 (Reference 13), February 25, 2014 (Reference 14), May 2, 2014 (Reference 15),
May 11, 2014 (Reference 16), August 14, 2014 (Reference 17), October 9, 2014 (Reference 18),
and December 11, 2014 (Reference 19), the licensee submitted an application for a license amendment to transition the VCSNS fire protection program 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 supplemental letters were in response to the NRC staff's requests for additional information dated July 26, 2012 (Reference 20),
August 13, 2013 (Reference 21), August 28, 2013 (Reference 22), November 7, 2013 (Reference 23), and December 19, 2013 (Reference 24), March 24, 2014 (Reference 25), and July 11, 2014 (Reference 26). The licensee's supplemental letters dated October 10, 2012; February 1, April 1, October 14, and November 26, 2013; January 9, February 25, May 2, May 11, August 14, October 9, and December 11, 2014, 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 staff's original proposed opportunity for a hearing on the initial application as published in the Federal Register on August 14, 2012 (77 FR 48561 ).
The licensee requested an amendment to the VCSNS renewed operating license and Technical Specifications (TSs) in order to establish and maintain a risk-informed, performance-based fire protection program in accordance with the requirements of 10 CFR 50.48(c).
Specifically, the licensee requested to transition from the existing deterministic fire protection licensing basis established in accordance with all provisions of the approved fire protection program as described in the VCSNS Final Safety Analysis Report and as approved in the Safety Evaluation Report dated February 1981 (Reference 27) and supplements dated January 1982 (Reference 28), and August 1982 (Reference 29), and Safety Evaluations dated May 22, 1986 (Reference 30), November 26, 1986 (Reference 31 ), and July 27 -1987 (Reference 32) to a risk-informed, performance-based fire protection program 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 fire protection program at VCSNS is referred to as risk-informed, performance-based throughout this SE.
In its license amendment request (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 fire protection program 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 fire protection program, as described in the LAR, in accordance with the implementation schedule set forth in this SE and the aGcompanying license condition, as required by 10 CFR 50.48(c)(3)(ii).
The licensee proposed a new fire protection license condition reflecting the new risk-informed, performance-based fire protection program licensing basis, as well as revisions to the Technical Specifications that address this change to the current fire protection program licensing basis.
Section 2.4.2 and Section 4.0 of this SE discuss in detail the license condition, and Section 2.4.3 discusses the TS changes.
2.0 REGULATORY EVALUATION
Section 50.48, "Fire Protection," of 10 CFR provides the NRC requirements for nuclear power plant fire protection. Section 50.48 includes specific requirements for requesting approval for a risk-informed, performance-based fire protection program based on the provisions of NFPA 805 (Reference 1 ). Paragraph 50.48(c)(3)(i) of 10 CFR states, in part, that:
A licensee may maintain a fire protection program that complies with NFPA 805 as an alternative to complying with paragraph (b) of this section [10 CFR 50.48(b)] for plants licensed to operate before January 1, 1979, or the fire protection license conditions for plants 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 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 fire protection program 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 FR 33536, 33548 (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 fire protection program 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 th.at 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 performance-based 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 an LAR in accordance with 10 CFR 50.48(c)(2)(vii). This regulation further provides that:
The Director of NRR, 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 (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 risk-informed or performance-based alternatives to comply with NFPA 805 by submitting a LAR in accordance with 10 CFR 50.48(c)(4). This regulation further provides that:
The Director of NRR, or designee of the Director, may approve the application if the Director or designee determines that the proposed alternatives:
(A)
Satisfy the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; (B)
Maintain safety margins; and (C)
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 risk-informed, performance-based fire protection program, a licensee may also submit additional elements of its fire protection program 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 2). Inclusion of these elements in the NFPA 805 LAR is meant to alleviate uncertainty in portions of the current fire protection program licensing bases as a result of the lack of specific NRC approval of these elements. RGs are not substitutes for regulations, and compliance with them is not required. Methods and solutions that differ from those set forth in RGs will be deemed acceptable if they provide a basis for the findings required for the issuance or continuance of a permit or license by the Commission. Accordingly, any submittal addressing these additional fire protection program elements needs to include sufficient detail to allow the NRC staff to assess whether the licensee's treatment of these elements meets the 10 CFR 50.48(c) requirements
- The purpose of the fire protection program 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 (3)
Providing an adequate level of fire protection for SSCs 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 General Design Criterion (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:
Structures, systems, and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.
10 CFR 50.48(a)(1 ), requires that each holder of an operating license have a fire protection plan that satisfies GDC 3 of Appendix A to 10 CFR Part 50.
10 CFR 50.48(c), incorporates NFPA 805 (2001 Edition) (Reference 1) by reference, with certain exceptions, modifications and supplementation. This regulation establishes the requirements for using a risk-informed, performance-based fire protection program in conformance with NFPA 805 as an.
alternative to the requirements associated with 10 CFR 50.48(b) and AppendixR, "Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979," to 10 CFR Part 50, or the specific plant fire protection license condition for plants licensed to operate after January 1, 1979.
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."
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 2), which provides guidance for use in complying with the requirements that the NRC has promulgated for risk-informed, performance-based fire protection programs 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 (Reference 5), whe~e 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 Reactor Electric Generating Plants," (Reference 1 ), 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 provide a comprehensive risk-informed, performance-based standard for fire protection.
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)," (Reference 5), which provides guidance for implementing the requirements of 10 CFR 50.48(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 incorporates guidance from RG 1.205 and approved Frequently Asked Questions (FAQs).
NEI 00-01, "Guidance for Post Fire Safe Shutdown Circuit Analysis," Revision 2, (Reference 33), which provides a deterministic methodology for performing post-fire safe shutdown analysis (SSA). In addition, NEI 00-01 includes information on risk-informed 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.
RG 1.17 4, "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 34), 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 1.17 4 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 35), 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
- 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.
American Society of Mechanical Engineers/American Nuclear Society (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 36), 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.
RG 1.189, "Fire Protection for Operating Nuclear Power Plants," Revision 2, issued October 2009 (Reference 37), which provides guidance to licensees on the proper content and quality of engineering equivalency evaluations used to support the fire protection program. 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 NRC 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 38), which provides guidance for NRC staff evaluations of LARs that seek to implement a risk-informed, performance-based fire protection program in accordance with 10 CFR 50.48(c).
NUREG-0800, Section 19.1, "Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 3, issued September 2012 (Reference 39), which provides guidance for NRC staff evaluations of the technical adequacy of a licensee's PRA results when used to request risk-informed 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 40), which provides guidance for NRC staff evaluations of the risk information used by a licensee to support permanent, risk-informed changes to the licensing basis for the plant.
NUREG/CR-6850, "EPRl/NRC-RES Fire PRA Methodology for Nuclear Power Facilities," Volumes 1 and 2, and Supplement 1, (References 41, 42, and 43),
which presents a compendium of methods, data and tools to perform a fire PRA 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.
Memorandum from Richard P. Correia, RES, to Joseph G. Giitter, NRR, titled "Interim Technical Guidance on Fire-Induced Circuit Failure Mode Likelihood Analysis," dated June 14, 2013, (Reference 44) notes that, based on new experimental information documented in NUREG/CR-6931 "Cable Response to Live Fire (CAROLFIRE)" issued April 2008 (Reference 45), and NUREG/CR-7100 "Direct Current Electrical Shorting in Response to Exposure Fire (DESIREE-Fire): Test Results," issued April 2012 (Reference 46), the reduction in hot short probabilities for circuits provided with control power transformers (CPTs) 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 47) 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.
NUREG-1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications," Volumes 1 through 7, (Reference 48), 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)
Fire Induced Vulnerability Evaluation (FIVE)-Rev1 developed by EPRI (Volume 4);
- 3)
The zone model consolidated model of fire and smoke transport (CFAST) developed by NIST (Volume 5);
- 4)
The zone model MAGIC developed by Electricite de France (EdF) (Volume 6); and
- 5)
The computational fluid dynamics model fire dynamics simulator (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 (CHRISTIFIRE), Volume 1: Horizontal Trays," (Reference 49), 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. CHRISTIFIRE 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 50), 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.
NUREG-1921, "EPRl/NRC-RES Fire Human Reliability Analysis Guidelines -
Final Report," (Reference 51 ), which presents the state of the art in fire human reliability analysis (HRA) practice. This report was developed jointly between RES and EPRI to develop the methodology and supporting guidelines for estimating human error probabilities for human failure events following the fire-induced initiating events of a fire PRA. The report builds on existing human reliability analysis methods, and is intended primarily for practitioners conducting a Fire HRA to support a Fire PRA.
Generic Letter {GL) 2006-03, "Potentially Nonconforming Hemyc and MT Fire Barrier Configurations" (Reference 51A), 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. Specifically, NRC testing revealed that, for the configurations tested, Hemyc and MT fire barriers failed to provide the protective function intended for compliance with existing regulations.
NFPA 101, "Life Safety Code" (Reference 67), provides the minimum requirements for egress, features of fire protection, sprinkler systems, alarms, emergency lighting, smoke barriers, and special hazard protection.
- *14 -
2.3 NFPA 805 Frequently Asked Questions In the LAR, the licensee proposed to use a number of documents commonly known as NFPA 805 Frequently Asked Questions (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.
FAQ#
07-0030 07-0038 Table 2 3-1 NFPA 805 Frequently Asked Questions FAQ Title and Summary Reference "Establishing Recovery Actions" 52 This FAQ provides an acceptable process for determining the 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.
"Lessons Learned on Multiple Spurious Operations (MSOs)"
53 This FAQ reflects an acceptable process for the treatment of MSOs during transition to NFPA 805:
Step 1 - Identify potential MSO combinations of concern.
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" 54 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.
SE Section 3.2.2 3.2.5 3.4.4 3.2.4 3.2.1 FAQ#
FAQ Title and Summary Reference SE Section 07-0040 "Non-Power Operations (NPO) Clarifications" 56 3.5.3
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.
08-0046 "Incipient Fire Detection Systems" 57 3.1.1.1
- This FAQ provides guidance for modeling 3.2.6 non-suppression probability when an incipient fire 3.4.2.2 detection system is installed to monitor electrical cabinets.
Applies to Aspirating Smoke Detectors (ASD) installed as Very Early Warning Fire Detectors (VEWFDS) to monitor incipient degradation in electrical cabinets.
08-0048 "Revised Fire Ignition Frequencies" 58 3.4.7
- 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.
08-0054 "Compliance with Chapter 4 of NFPA 805" 59 3.4.3 This FAQ provides an acceptable process to 3.5.1.4 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 and Features 09-0056 "Radioactive Release Transition" 60 3.6.1 This FAQ provides 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.
\\ FAQ#
FAQ Title and Summary Reference SE Section 10-0059 "Monitoring Program" 61 3.7 This FAQ provides clarification regarding the implementation of an NFPA-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.
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 LAR Section 5.2.3, "Orders and Exemptions" and LAR Attachment 0, "Orders and Exemptions" of the LAR, with regard to NRG-issued Orders that are being revised or superseded by the NFPA 805 transition process. The LAR stated that the licensee conducted a review of 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 are maintained. The licensee discussed the affected orders and exemptions in LAR Attachment 0.
Since VCSNS is a post-1979 operating plant, 10 CFR 50, Appendix R, exemptions were not necessary and are therefore not applicable to VCSNS. The licensee determined that no Orders need to be superseded or revised to implement a fire protection program that complies with 10 CFR 50.48(c).
The review conducted by SCE&G, included an assessment of docketed correspondence files and electronic searches, including the NRC's 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 VCSNS are maintained. The NRC staff accepts the licensee's determination that no exemptions are rescinded and that no Orders need to be superseded or revised to implement NFPA 805 at VCSNS. See SE Section 2.5 for further discussion.
In addition, the licensee performed a specific review of the license amendment that incorporated the mitigation strategies required by Section 8.5.b of Commission Order EA-02-026 (subsequently incorporated into 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 VCSNS. The licensee's review of this order and the related license amendment demonstrated that changes to the fire protection program during transition to NFPA 805 will not affect the mitigation measures required by Commission Order EA-02-026. The NRC staff concludes that the licensee's determination in regard to Commission Order EA-02-026 is acceptable.
2.4.2 License Conditions The NRC staff reviewed LAR Section 5.2.1, "License Condition Changes," and LAR Attachment M, "License Condition Changes," regarding changes the licensee seeks to make to the VCSNS fire protection license condition in order to adopt NFPA 805, as required by 10 CFR 50.48(c)(3).
The NRC staff reviewed the revised license condition, which supersedes the current VCSNS fire protection license condition, 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 revised license conditions provide a structure and detailed criteria to allow self-approval for risk-informed, performance-based as well as other types of changes to the fire protection program. The structure and detailed criteria result in a process that meets the requirements in NFPA 805 Sections 2.4, Engineering Analyses; 2.4.3, Fire Risk Evaluations; and 2.4.4, Plant Change Evaluation. These sections establish the requirements for the content and quality of the engineering evaluations to be used for approval of changes.
The revised license conditions also define 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 licensee's revised license condition provides structure and detailed criteria to allow self-approval for fire protection program changes that meet the requirements of NFPA 805 with regard to Engineering Analyses, Fire Risk Evaluations and Plant Change Evaluations. The NRC staff's evaluation of the self-approval process for fire protection program changes (post-transition) is contained in SE Section 2.6. The license condition also references the plant-specific modifications, implementation items, and associated schedules that must be met at VCSNS to complete transition to NFPA 805 and comply with 10 CFR 50.48(c). In addition, the license condition includes a requirement that appropriate compensatory measures will remain in place until completion of the specified plant modifications. These modifications, implementation items, and schedules are identical to those identified elsewhere in the LAR, as discussed by the NRC staff in SE Section 2.7.
SE Section 4.0 provides the NRC staff's review of the VCSNS fire protection program license condition.
2.4.3 Technical Specifications The NRC staff reviewed LAR Section 5.2.2, "Technical Specifications" and LAR Attachment N,
Technical Specification Changes," with regard to proposed changes to the VCSNS Technical Specifications (TSs) that are being revised or superseded during the NFPA 805 transition process. According to the LAR, the licensee conducted a review of the VCSNS TSs to determine which, if any, TS sections will be impacted by the transition to a risk-informed, performance-based fire protection program based on 10 CFR 50.48(c). The licensee identified changes to the Technical Specifications needed for VCSNS adoption of the new fire protection licensing basis and provided applicable justification listed in LAR Attachment N. The licensee identified one change to the TS that involved deleting TS 6.8.1.f, which requires that procedures be established, implemented, and maintained for the fire protection program.
Specifically, the licensee stated that deleting TS 6.8.1.f is acceptable for adoption of the new fire protection licensing basis because the requirement for establishing, implementing, and maintaining fire protection procedures is contained in 10 CFR 50.48(a) and 10 CFR 50.48(c) and because 10 CFR 50.48(b) Appendix R requirements will be superseded by 10 CFR 50.48(a) and 10 CFR 50.48(c). 10 CFR 50.48(c) approves the incorporation of NFPA 805 by reference and NFPA 805 Section 3.2.2, "Procedures," states that "Procedures shall be established for implementation of the fire protection program."
Based on the information provided by the licensee, the NRC staff concludes that the proposed change to the TS is acceptable because TS 6.8.1.f is an administrative control (i.e., a procedure the licensee puts in place to establish, implement, and maintain the fire protection program as required by the licensee's fire protection license condition and 10 CFR 50.48(a),
10 CFR 50.48(c), and NFPA 805, Section 3.2.3), and would be redundant to the NFPA 805 requirement to establish fire protection program procedures. NFPA 805 requires the licensee to establish fire protection program procedures, and 10 CFR 50.48(a) and 10 CFR 50.48(c) would become the fire protection licensing basis of VCSNS. In addition, failure by the licensee to establish fire protection program procedures would result in non-compliance with 10 CFR 50.48(c)(1 ), which is the licensee's fire protection licensing basis. Changes to fire protection ad.ministrative controls are controlled by the proposed fire protection license condition.
See SE Section 4.0.
2.4.4 Updated Final Safety Analysis Report' The NRC staff reviewed the LAR and noted that Figure 4-8, "NFPA 805 Planned Post-Transition Documents and Relationships," of the LAR indicates that post-transition NFPA 805 documentation will include the revised license condition and UFSAR.
Updates to the UFSAR are required by 10 CFR 50.71 (e), and the licensee stated in a letter dated November 14, 1980 (Reference 62) that records are and will be maintained in accordance with the requirements of sections (a) through (e) of the regulation (10 CFR 50.71) and the license.
Since the licensee's process for updating its UFSAR is in accordance with 10 CFR 50.71(e),
which is consistent with the guidance provided in NEI 04-02 for updating the UFSAR, the NRC staff concludes that the licensee's method for updating the UFSAR is acceptable.
2.5 Rescission of Exemptions The NRC staff reviewed LAR Section 5.2.3, "Orders and Exemptions," LAR Attachment 0, "Orders and Exemptions," and LAR 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 fire protection program licensing basis in conformance with NFPA 805will supersede.
VCSNS was licensed to operate after January 1, 1979 and therefore licensing actions associated with 10 CFR 50 Appendix R were not issued as exemptions to the regulation. Therefore no exemptions need to be rescinded. Additionally the licensee stated that a review of VCSNS docketed correspondence to determine if any Orders needed to be superseded or revised was performed. No Orders were found requiring revision or superseding.
2.6 Self Approval Process for Fire Protection Program Changes (Post-Transition)
Upon completion of the implementation of the risk-informed, performance-based fire protection program and issuance of the license conditions discussed in SE Section 2.4.2, changes to the approved fire protection program must be evaluated by the 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 risk-informed plant change evaluation shall be performed and the results used as described in 2.4.4 to ensure that the public risk associated with fire-induced nuclear fuel damage accidents is low and that adequate defense-in-depth and safety margins are maintained.
NFPA 805, Section 2.4.4, "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, "NFPA 805 Configuration Management (NFPA 805, Sections 2.2.9/2.7.2)," for compliance with the NFPA 805 plant change evaluation process requirements to address potential changes to the NFPA 805 risk-informed, performance-based fire protection program after-implementation is completed. The licensee developed a change process that is based on the guidance provided in NEI 04-02, Revision 2 (Reference 5), Section 5.3, "Plant Change Process," as well as Appendices B, I, and J, as modified by RG 1.205, Revision 1, (Reference 2), Regulatory Positions 2.2.4, 3.1, 3.2, and 4.3.
LAR Section 4. 7.2 states that the Plant Change Evaluation (PCE) process consists of four steps:
- 1)
Defining/Screening the change;
- 2)
Performing the Preliminary Risk Screening;
- 3)
Performing the Risk Evaluation; and
- 4)
Evaluating the Acceptance Criteria.
In the LAR, the licensee stated that the plant change evaluation process begins by defining the change or altered condition 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 is then performed to identify and resolve minor changes to the fire protection program and that the screening is consistent with fire protection regulatory review processes currently in place at nuclear plants under traditional licensing bases. The licensee further stated that the screening process i~ modeled after NEI 02-03, "Guidance for Performing a Regulatory Review of Proposed Changes to the Approved Fire Protection Program," June 2003, (Reference 63) and that the process will address most administrative changes (e.g., changes to the combustible control program, organizational changes, etc.).
The licensee stated that once the screening process is completed, it is followed by engineering evaluations that might include fire modeling and risk assessment techniques and that the results of these evaluations are then 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 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 involves 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 change evaluations are assessed for acceptability using the delta (!l)
CDF (change in CDF) and~ LERF (change in LERF) criteria from the license condition and that the proposed changes are also assessed to ensure they are consistent with the defense-in-depth philosophy and that sufficient safety margins were maintained.
The licensee stated that its fire protection program configuration is defined by the progra~
documentation and, to the greatest extent possible, the existing configuration control processes for modifications, calculations and analyses. The licensee further stated that fire protection program license basis reviews will be utilized to maintain configuration control of the fire protection program documents. The licensee furt_her stated that the configuration control procedures that govern the various VCSNS documents and databases that currently exist will be
- revised to reflect the new NFPA 805 licensing bases requirements. This action is included in LAR Attachment S, Table S-2, Implementation Items 13 and 14 and is considered acceptable because it is included in the proposed license condition and will be consistent with the configuration control processes currently in place.
The licensee stated that several NFPA 805 document types such as: nuclear safety capability assessment (NSCA) Supporting Information, Non-Power Operations Treatment, 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. The licensee further stated that the new processes 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. This action is included in LAR Attachment S, Table S-2, Implementation Item 18 and is considered acceptable because it is included in the proposed license condition and will be consistent with the configuration control processes currently in place.
The licensee stated that the process for capturing the impact of proposed changes to the plant on the fire protection program will continue to be a multiple step review and that the first step of the review is an initial screening for process users to determine if there is a potential to impact the fire protection program 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 fire protection program impacts will be sent to qualified individuals (e.g., Fire Protection, Safe/Shutdown/NSCA, Fire PRA, as applicable) to ascertain the program impacts, if any, and that if fire protection program impacts are determined to exist as a result of the proposed change, the issue would be resolved by one of the following:
Deterministic Approach: Comply with NFPA 805 Chapter 3 and 4.2.3
- requirements.
Performance-Based Approach: Utilizing the NFPA 805 change process developed in accordance with NEI 04-02, RG 1.205, and the VCSNS NFPA 805 fire protection license condition to assess the acceptability of the proposed change.
This process would be used to determine if the proposed change could be implemented as-is or whether 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 1.17 4, (Reference 34 ), which requires the use of qualified individuals, procedures that require calculations 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.
Since NFPA 805 always requires the use of a plant change evaluation (PCE), regardless of what element requires the change, the NRC staff concludes that, in accordance with the requirements of NFPA 805, iffire protection program impacts are determined to exist as a result of the proposed change, the issue would be resolved by utilizing the NFPA 805 change process developed in accordance with NEI 04-02, RG 1.205, and the VCSNS 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.
Based on the information provided by the licensee, the NRC staff concludes that the licensee's plant change evaluation process is acceptable because it meets the guidance in NEI 04-02, Revision 2, (Reference 5), as well as RG 1.205, Revision 1, (Reference 2), 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 appro~ch, 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, defense-in-depth, 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 utilize methods that are acceptable to the NRC staff. Acceptable methods to assess the risk of the proposed plant change 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 9mendment 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 5), as well as RG 1.205, Revision 1, (Reference 2). The NRC staff concludes that the proposed plant change evaluation process at VCSNS, which includes defining the change, a preliminary risk screening, a risk evaluation, and an acceptability determination, as described in SE 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, defense-in-depth, 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), risk-informed changes to the licensee's fire protection program 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 defense-in-depth 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 2), with respect to the requirements for fire protection program 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 fire protection program that may be made on a qualitative, rather than quantitative basis.
Specifically, the license conditions state that 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 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 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 or 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 allow~d by NFPA 805 Section 2.4.
The four 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 3.1 O); 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. 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 risk-informed changes to the fire protection program under the proposed license conditions, the NRC staff concludes that the licensee's process for self-approving future fire protection program changes are acceptable.
The NRC staff also concludes that the fire risk evaluation methods used at VCSNS 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 risk-informed, performance-based fire protection program, based on the licensee's administrative controls to ensure that the models remain current and to assure continued quality (see SE Section 3.4.1, "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 fire risk evaluations conducted to determine the change in risk associated with proposed plant changes.
- 2. 7 Modifications and Implementation Items Regulatory Position C.3.1 of RG 1.205, Revision 1, (Reference 2), 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.
The list of modifications and implementation items originally submitted in the LAR have been updated by the licensee with the final version of LAR Attachment S, "Plant Modifications and Items to be Completed during Implementation." The updated LAR Attachment Sis provided in the licensee's letter dated December 11, 2014 (Reference 19).
2.7.1 Modifications The NRC staff reviewed LAR Attachment S, "Plant Modifications and Items to be Completed During Implementation," which describes the plant modifications necessary to implement the NFPA 805 licensing basis, as proposed. These modifications are identified in the LAR as necessary to bring VCSNS into compliance with either the deterministic or performance-based requirements of NFPA 805. As described below, LAR Attachment S, Table S-1, 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 Attachment S, 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 Attachment S, Table S-1 modifications and associated completion schedule are the same as those provided in the proposed NFPA 805 license condition.
As depicted in LAR Attachment S, Table S-3, the licensee has completed 5 modifications as part of the NFPA 805 transition. LAR Attachment S, Table S-1, provides a detailed listing of the plant modifications that must be completed in order for VCSNS to be fully in accordance with NFPA 805, implement many of the attributes upon which this SE is based, and thereby meet the requirements of 10 CFR 50.48(c). The modifications will be completed in accordance with the schedule provided in the proposed NFPA 805 license condition, which states that all modifications will be in place by the end of calendar year 2015. *In addition, the licensee stated that it will keep the appropriate compensatory measures in place until the modifications described in LAR Attachment S, Table S-1, have been completed, as stated in the proposed license condition.
2.7.2 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). The licensee identified the implementation items in LAR Attachment S, Table S-2. 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. Completion of these items in accordance with the schedule discussed in SE Section 2.7.3, does not change or impact the bases for the safety conclusions made by the NRC staff in the SE.
Each implementation item will be completed prior to the deadline for implementation of the risk-informed, performance-based fire protection program based on NFPA 805, as specified in the license condition and the letter transmitting the amended license (i.e., implementation period) which states that the implementation items listed in LAR Attachment S, Table S-2, will be completed by March 31, 2016.
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.
2.7.3 Schedule LAR Section 5.4 provides the overall schedule for completing the NFPA 805 transition at VCSNS.
The licensee stated that it will complete the implementation of the new program, including any procedure changes, process updates, and training for affected plant personnel to implement the NFPA 805 fire protection program within 180 days after NRC approval except for the items listed in b. and c. below, that will take longer to complete. In a letter dated December 11, 2014 (Reference 19), the licensee provided a revised schedule for implementing the implementation items listed in LAR Attachment S, Table S-2, as follows:
- a.
- items 3, 6, 7, 8, 10, 13, 14, 17, 19, and 21 will be implemented within 180 days of NRC approval.
- b.
Items 1, 2, 4, 11, and 12 will be implemented by December 31, 2015.
- c.
Items 5, 15, 16, 18, 20, 22, and 23 will be implemented by March 31, 2016.
LAR Section 5.4 also states that the modification schedule is provided in LAR Attachment S. LAR Attachment S, Table S-1, indicates that all modifications will be completed by the end of calendar year 2015.
3.0 TECHNICAL EVALUATION
The following sections evaluate the technical aspects of the requested license amendment to transition the fire protection program at VCSNS to one based on NFPA 805 (Reference 1) in accordance with 10 CFR 50.48(c). While performing the technical evaluation of the licensee's submittal, the NRC staff utilized the guidance provided in NUREG-0800, Section 9.5.1.2, "Risk-Informed, Performance-Based Fire Protection" (Reference 38), 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 SE Section 2.0. Specifically:
Section 3.1 provides the results of the NRC staff review of the licensee's transition of the fire protection program from the existing deterministic guidance to that of NFPA 805 Chapter 3, "Fundamental Fire Protection Program 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.
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 nuclear safety performance criteria using a fire modeling performance-based 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 nuclear safety performance criteria using a fire risk evaluation performance-based approach.
Section 3.5 provides the results of the NRC staff review of the licensee's 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 risk-informed, performance-based fire protection program 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, SE Attachments A and B provide additional detailed information that was evaluated by the NRC staff to support the licensee's request to transition to a risk-informed, performance-based fire protection program in accordance with NFPA 805 (i.e., 10 CFR 50.48(c)).
These attachments are discussed as appropriate in the associated SE sections.
3.1.
NFPA 805 Fundamental Fire Protection Program (FPP} and Design Elements NFPA 805 (Reference 1) Chapter 3 contains the fundamental elements of the fire protection program and specifies the minimum design requirements for fire protection systems and features that are necessary to meet the standard. The fundamental fire protection program elements and minimum design requirements include necessary attributes pertaining to the fire protection plan and procedures, the fire prevention program and design controls, 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.
1 O 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 3.6A 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 performance-based methods to demonstrate compliance with the NFPA 805, Chapter 3 requirements, 10 CFR 50.48(c)(2)(vii) specifically permits that the fire protection program elements and minimum design requirements of NFPA 805, Chapter 3 may be subject to the performance-based methods permitted elsewhere in the standard, provided a license amendment is granted and the approach satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release, maintains safety margins; and maintains fire protection defense-in-depth.
Furthermore, Section 3.1 of NFPA 805 specifically allows the use of alternatives to the NFPA 805, Chapter 3 fundamental fire protection program requirements that have been previously approved by the NRC (which is the authority having jurisdiction (AHJ), as denoted in NFPA 805 and RG 1.205), and are contained in the currently approved fire protection program 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 5), as endorsed by the NRC in Regulatory Guide 1.205, Revision 1 (Reference 2), to assess the proposed VCSNS fire protection program 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 VCSNS fire protection program and provided specific compliance statements for each NFPA 805, 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 VCSNS, and others are provided with multiple compliance statements to fully document compliance with the element.
The methods used by VCSNS for achieving compliance with the fundamental fire protection program elements and minimum design requirements are as follows:
- 1.
The existing fire protection program element directly complies with the requirement: noted in LAR Attachment A, "NEI 04-02 Table B-1, Transition of Fundamental Fire Protection Program and Design Elements," (also called the B-1 Table), as "Complies (C)." (see SE Section 3.1.1.1)
- 2.
The existing fire protection program elements meet the requirements of NFPA 805 by using clarification and/or equivalent alternative(s). VCSNS requests NRC review/approval of those complies by alternative items listed in LAR Section 4.1.2.3 (Table 4-1) and included in Attachment L: noted LAR Table 8-1 as "Complies by Alternative (CA)". (see SE Section 3.1.1.2)
- 3.
The existing fire protection program element complies through the use of Fire Protection Engineering Equivalency Evaluations (FPEEEs) whose basis remains valid and is of sufficient quality: noted in the 8-1 Table as "Complies with Fire Protection Engineering Equivalency Evaluations (CE)". (see SE Section 3.1.1.3)
- 4.
The existing fire protection program element complies with the requirement based on prior NRC approval of an alternative to the fundamental fire protection program attribute and the bases for the NRC approval remain valid: noted in the B-1 Table as "Complies by Previous NRC Approval (CNRC)." (see SE Section 3.1.1.4)
- 5.
The existing fire protection program element does not comply with the requirement, but the licensee is requesting specific approval for a PB method in accordance with 10 CFR 50.48(c)(2)(vii): noted in LAR Table B-1 as "Submit for NRC Approval." (see SE Section 3.1.1.5)
- 6.
The existing NFPA 805 Chapter 3 elements are not based on the requirements and/or are not applicable to elements of the VCSNS Fire Protection Program:
noted in the B-1 Table as "No Review Required (NRR)." (see SE Section 3.1.1.6)
Compliance approach 2, "Complies by Alternative (CA)," is a modification from the NEI 04-02 based approach in that it is a new category not included in NEI 04-02. The licensee combined the NEI 04-02 categories of compliance; "Complies with Clarification", and "License Amendment Required" and referred to the compliance approach as "Complies by Alternative (CA)", where the existing FPP elements meet the requirements of NFPA 805 by using clarification and/or equivalent alternative(s). The guidance provided in NEI 04-02 indicates that the "Complies with Clarification" category is for items that are not in "literal compliance" with the requirement as listed in NFPA 805 but should be transitioned as complies.
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 in that the "Complies by Alternative" category used by the licensee includes approaches that either require clarification or a license amendment, which are both acceptable approaches per the guidance of NEI 04-02.
The licensee stated in LAR Section 4.2.2, "Fire Protection Engineering Equivalency Evaluation Transition," that it evaluated the FPEEEs used to demonstrate compliance with the NFPA 805, Chapter 3 or 4 requirements to'ensure continued appropriateness, quality, and applicability to the current VCSNS plant configuration. The licensee stated that any FPEEEs where SCE&G requested specific NRC approval are included in LAR Attachment Kor LAR Attachment L.
In a FPE RAI 14 (Reference 20) the NRC staff requested that the licensee specifically identify which elements of LAR Table B-1, that utilized the "Complies by Alternative" compliance approach, required NRC staff approval. In its response to FPE RAI 14 (Reference 8) the licensee identified the following elements as requiring NRC staff approval:
3.3.1.2(1) 3.4.1 (d) 3.4.2.4 3.4.3(a)(4) 3.6.2 3.3.5.1 3.3.5.3 Each of the above elements, in addition to elements that utilize other compliance approaches such as elements 3.3.7.2 and 3.5.15, are already identified in LAR Attachment Las requiring NRC Staff approval as 10 CFR 50.48 (c)(2)(vii) performance-based alternative compliance strategies and therefore are evaluated in SE Section 3.1.4.
EEEEs refer to "existing engineering equivalency evaluations" (previously known as Generic Letter 86-10 evaluations as defined in NEI 04-02) performed for fire protection design variances such as fire protection system designs and fire barrier component deviations from the specific fire protection deterministic requirements. VCSNS has chosen to label these as "Fire Protection Engineering Equivalency Evaluations" (FPEEEs) in the LAR. 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 Actions: Resulting NFPA 805 Analysis & Transitions," 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 LAR Attachment K.
LAR Attachment A (the NEI 04-02 Table 8-1) 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 the majority of the NFPA 805 Chapter 3 requirements, as modified by 10 CFR 50.48(c)(2), the licensee determined that the risk-informed, performance-based fire protection program complies directly with the fundamental fire protection program element using the existing fire protection program element. In these instances, based on the information provided by the licensee in the LAR, the NRC staff concludes that the licensee's statements of compliance are acceptable.
The following NFPA 805 sections identified in LAR Table 8-1 as complying via this method required additional review by the NRC staff:
3.8.1 3.11.2 3.3.9 3.11.3 NFPA 805, Section 3.8.1, "Fire Alarm System" requires that alarm initiating devices be installed in accordance with NFPA 72, "National Fire Alarm Code" (Reference 64). Incipient detection (Very Early Warning Fire Detection System) is described as a necessary modification in LAR Attachment S, Table S-1, Item ECR50811 for certain electrical cabinets in the Relay and Upper Cable Spread Rooms. In FPE RAI 01 (Reference 20) the NRC staff requested additional information regarding the proposed modification for a VEWFDS. In its response to FPE RAI 01 (Reference 8), the licensee stated the following:
FAQ 08-00046 (Reference 57), Electric Power Research Institute (EPRI)
Technical Report (TR) 1006756, "Fire Protection Equipment Surveillance Optimization and Maintenance Guide," Final Report, July 2003 (Reference 65),
and NUREG-6850, Supplement 1 (Reference 43) will be used as design inputs in system selection, design, and installation. This includes design features, acceptance testing, and routine inspection testing and maintenance.
In accordance with FAQ 08-0046, NFPA 76 (Reference 66), "Standard for the Fire Protection of Telecommunications Facilities" will be used as the code of record for the design of the system at VCSNS.
A code compliance document will be developed and maintained by VCSNS after installation.
Sensitivity settings will meet those defined in the November 23, 2009, FAQ 08-0046 Closure Memorandum.
Regular Functional Testing and Maintenance will be performed in accordance with NFPA 76, NFPA 72, and applicable vendor recommendations.
The system will be included within the overall NFPA 805 condition monitoring program.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated that the installation of the VEWFDS will meet the intent of NFPA 805, Section 3.8.1 since the licensee will utilize staff accepted guidance and national consensus standards in the design, installation and testing.
NFPA 805 Section 3.3.9 "Transformers" requires that transformer oil collection basins and drain paths be periodically inspected to ensure they are free of debris and. capable of performing their design function. LAR Attachment A, Section 3.3.9 indicated that a fire hazard evaluation of the transformer area considered drainage alternatives to that cited in this section. In FPE RAI 06 (Reference 20) the NRC staff requested that the licensee provide clarification of the alternatives used. In its response to FPE RAI 06 (Reference 8), the licensee indicated that "VCSNS relies upon a function test in lieu of a periodic inspection to ensure that the oil collection basins are capable of performing their design function. During Transformer Deluge System Testing, flow is maintained for a 10 minute water spray discharge plus an added duration based on the oil volume of the subject transformer. The test confirms that deluge water is contained in the transformer catch basin collection system." The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated that the alternative provided for the transformer oil collection basins and drain paths meets the intent of NFPA 805, Section 3.3.9 through the performance of a functional test.
NFPA 805, Section 3.11.2 "Fire Barriers" requires that fire barriers, required by NFPA 805 Chapter 4, include a specific fire resistance rating. In FPE RAI 17 (Reference 20) the NRC staff requested that the licensee provide confirmation that no active fire barriers (such as water curtains), thermal wraps, electric raceway fire barrier systems (ERFBS), or fire resistant coatings are used to credit partitioning boundaries in the Fire PRA. In its response to NRC FPE RAI 17 (Reference 8), the licensee stated that no active fire barriers, thermal wraps, ERFBS or fire
- resistant coatings are credited as partitioning elements in the Fire PRA and that no active fire barrier elements, thermal wraps or fire resistant coatings are credited in the plant regulatory fire protection program. In LAR Attachment A, Table B-1 Section 3.11.5 the licensee stated that the ERFBS that are required by Chapter 4 have a 1-hour or 3-hour fire rating and have been tested in accordance with the requirements of NFPA 805 Section 3.11.5. The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee did not credit any active fire barriers, therrnal wraps, fire resistant coatings or ERFBS as partitioning boundaries in the Fire PRA, and ERFBS that are required by Chapter 4 are rated and tested in accordance with NFPA 805 requirements, therefore the plant fire barriers meet the intent of NFPA 805, Section *3.11.2.
NFPA 805, Section 3.11.3 "Fire Barrier Penetrations," requires that for fire barrier penetrations the licensee comply with applicable standards including NFPA 101, Life Safety Code (Reference 67). In FPE RAI 18 (Reference 20), the NRC staff requested that the licensee identify the compliance strategy for penetrations in fire barriers. In its response* to FPE RAI 18 (Reference 8), the licensee stated that the code compliance document and Table B-1 would be revised to incorporate applicable sections of NFPA 101 related to maintaining the integrity of fire rated barriers via protection of personnel and equipment openings including fire protection devices such as doors and dampers. The NRC staff found that LAR Attachment S, Table S-2, includes Items 11 and 12 that address updating LAR Attachment A, Table B-1 as well as surveillance procedures and installation details to bring VCSNS into compliance with NFPA 805, Section 3.11. The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee identified a required action that will incorporate the provisions of NFPA 805, Chapter 3 in the licensee's fire protection program and included the action as an implementation item in LAR Attachment S, which would be required by the proposed license condition.
Based on the licensee's 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 completed during the implementation phase; the NRC staff concfudes 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 by Alternative For a small number of the NFPA 805, Chapter 3 requirements, the licensee provided additional clarification when describing its means of compliance with the fundamental fire protection program element. In these instances, the NRC staff reviewed the additional clarifications and concludes that the licensee will meet the underlying requirement for the fire protection program element as clarified.
In FPE RAI 14 (Reference 20), the NRC staff requested that the licensee provide a listing of each Chapter 3 requirement that uses the complies by alternative compliance strategy and to also describe whether the change is an editorial clarification or an actual request for approval. In its response to FPE RAI 14 (Reference 8), the licensee identified the requests for approval and the
- NRC staff has reviewed each in accordance with 10 CFR 50.48 (c)(2)(vii) performance-based alternative compliance strategies and therefore these elements are addressed in SE Section 3.1.4.
The following NFPA 805 sections identified in LAR Attachment A, Table B-1 as complying via clarification required additional review by the NRC staff:
Section 3.3.6 Roofs NFPA 805, Section 3.3.6 requires that roof coverings shall be Class A as determined by tests described in NFPA 256, "Standard Methods of Fire Test of Roof Coverings" (Reference 68). In FPE RAI 14 (Reference 20), the NRC staff requested that the licensee provide a listing of each Chapter 3 requirement that uses the complies by alternative compliance strategy and to also describe whether the change is an editorial clarification or an actual request for approval. In FPE RAI 14.01.c (Reference 21 ), the NRC staff requested that the licensee provide a detailed description of how metal roof construction, identified in Table B-1, Section 3.3.6, meets the requirements of Section 3.3.6 of NFPA 805. In its response to FPE RAI 14 (Reference 8) the licensee stated that metal roof construction is Class 1 per Factory Mutual Global Property Loss Prevention Data Sheet 1-31. In its response to FPE RAI 14.01.c (Reference 11 ), the licensee stated that NFPA 256 was not an original design requirement for the plant and that NRC Branch Technical Position (BTP) Auxiliary and Power Conversion Systems Branch (APCSB) 9.5-1 Appendix A was the prior licensing commitment for VCSNS. The licensee further stated that BTP APCSB 9.5-1 Appendix A, Position D.1.(e) required metal deck roof construction to be non-combustible in accordance with the Underwriters Laboratory (UL), Inc. or listed as Class I by Factory Mutual System Approval Guide. The licensee further stated that it complies with this requirement for metal deck roof construction as non-combustible and listed as Class I by the Factory Mutual System Approval Guide, which is documented in the Fire Protection Evaluation Report (FPER) May 2013, and that additional requirements for roof deck assemblies are provided by National [sic: Nuclear] Electric Insurance Limited (NEIL) in the NEIL Loss Control Manual, March 2011 Edition, Section 3.2.4.1. The licensee further stated that subsequent roof replacements involving metal deck construction required the continued compliance with Factory Mutual Class I roof assemblies as noted in the VCSNS construction specification SP-152, "Roof Insulation, Built Up Roofing and Sheet Metal," Revision 3 and that an engineering review (EIR 81843) was completed to confirm this requirement and installation. The NRC staff concludes that the licensee's responses to the RAls are acceptable because the licensee demonstrated metal deck roof construction meets the intent of NFPA 805, Section 3.3.6 by complying with the Class I non-combustible requirements in accordance with the Factory Mutual System Approval Guide.
3.1.1.3 Compliance Strategy -- Complies with Use of FPEEEs For certain NFPA 805, Chapter 3 requirements, the licensee demonstrated compliance with the.
fundamental fire protection program element through the use of FPEEEs. The NRC staff reviewed the licensee's statement of continued validity for the FPEEEs in LAR Section 4.2.2, as well as a statement on the quality and appropriateness of the evaluations, and finds the licensee's statements of compliance in these instances acceptable.
3.1.1.4 Compliance Strategy -- Complies via Previous NRC Approval Some NFPA 805, Chapter 3 requirements were supplanted by an alternative that was previously approved by the NRC. The NRC approval was documented in (1) the original February 1981 fire protection program Safety Evaluation Report (Reference 27), (2) and Supplements dated January 1982 and August 1982 (References 28 and 29).
Section 3.11.3(1) Fire Barrier Penetrations NFPA 80, Standard for Fire Doors and Fire Windows (Reference 69): Prior NRC approval of specialty doors (e.g.,
pressure and bullet resistant) was previously found to be acceptable and remain valid.
Section 3.11.3(2) Fire Barrier Penetrations NFPA 90A, Standard for Installation of Air-Conditioning and Ventilating Systems (Reference 70): Prior NRC approval of unique damper configurations was previously found to be acceptable and remain valid.
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 concludes that previous NRC approval had been demonstrated using suitable documentation that meets the approved guidance contained in RG 1.205, Revision 1 (Reference 2). 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 finds the licensee's statements of compliance in these instances acceptable.
3.1.1.5 Compliance Strategy - Submit for NRC Approval The licensee requested approval for the use of performance based methods to demonstrate compliance with fundamental fire protection program elements. In accordance with 10 CFR 50.48(c)(2)(vii), the licensee requested specific approvals be included in the license amendment approving the transition to NFPA 805. The NFPA 805 sections identified in LAR Attachment A, Table B-1 as complying via this method are as follows:
- 3.3.7.2
- 3.5.15 3.6.4 3.8.2 The licensee defined a compliance category, "compliance by alternative" or (CA) that included both compliance by "clarification" and compliance by "request for approval". In FPE RAI 14 (Reference 20), the NRC staff requested that the licensee provide a listing of each Chapter 3 requirement that uses the "complies by alternative" compliance strategy and to also describe whether the change is an editorial clarification or an actual request for approval. In its response to FPE RAI 14 (Ref~rence 8), the licensee also identified the following elements as requiring NRC staff approval:
3.3.1.2(1)
- 3.4.1 (d) 3.4.2.4 3.4.3.(a) (4)
- 3.6.2 3.3.5.1 3.3.5.3 NFPA 805, Section 3.3.1.2(1) "Control of Combustibles Materials" requires that wood used in the power block be listed pressure-impregnated or coated with a listed fire-retardant application. Contrary to this requirement, the licensee indicated that untreated wood or lumber will be addressed for controls of limited duration as part of the fire protection program using compensatory measures which are included in Implementation Item 1, in LAR Attachment S, Table S-2.
See SE Section 3.1.4.1 for the NRC staff's review of this request.
NFPA 805, Section 3.3.5.1 requires that wiring above suspended ceilings be kept to a minimum, and that where installed, electrical wiring shall be listed for plenum use, routed in armored cable, routed in metal conduit, or routed in cable trays with solid metal top and bottom covers. The licensee indicated that there is existing wiring for non-essential, non-risk significant areas and systems such as lighting and electrical power outlets that may not meet these requirements. See SE Section 3.1.4.2 for the NRC staff's review of this request.
NFPA 805, Section 3.3.5.3 requires that electric cable construction comply with a flame propagation test as acceptable to the Authority Having Jurisdiction. The
. licensee has indicated that very small amounts of untested special purpose cable are installed but do not result in a significant fire risk. See SE Section 3.1.4.3 for the NRC staff's review of this request.
NFPA 805, Section 3.3.7.2 requires that outdoor high-pressure flammable gas storage containers be located so that the long axis is not pointed at buildings. The substantial distance of the hydrogen storage tanks from the Turbine and Control buildings is an alternative approach to the prescriptive requirement of the code regarding the orientation of the tank axis. See SE Section 3.1.4.4 for the NRC staff's review of this request.
NFPA 805, Section 3.4.1 (d) requires that the fire brigade be notified immediately upon verification of a fire. The licensee indicated that the Control Room notifies the fire brigade upon verification of a fire. See SE Section 3.1.4.5 for the NRC staff's review of this request.
NFPA 805, Section 3.4.2.4 requires that pre-fire plans address coordination with other plant groups during fire emergencies. The licensee requests approval for the use of multiple procedures to coordinate the fire brigade activities with other groups. See SE Section 3.1.4.6 for the NRC staff's review of this request.
NFPA 805, Section 3.4.3(a)(4) requires that written records for fire brigade training be maintained for each industrial fire brigade member. The licensee has indicated that these records are kept electronically or written. See Section 3.1.4. 7 of this SE for the NRC staff's review of this request.
NFPA 805 Section 3.5.15 requires that hydrants be installed approximately every 250 ft apart on the yard main system. The licensee indicated that hydrants are spaced on average every 325 feet in the Protected Area. See SE Section 3.1.4.8 for the NRC staff's review of this request.
NFPA 805, Section 3.6.2 requires that a capability be provided to ensure adequate water flow rate and nozzle pressure for all hose stations including the provision of pressure reducers. The licensee indicated that pressure reducers have not been provided or are not necessary because the training of fire brigade members addresses high pressure nature of the system. See SE Section 3.1.4.9 for the NRC staff's review of this request.
NFPA 805 Section 3.6.4 requires that provisions be made to supply water at least to standpipes and hose stations for manual fire suppression in all areas containing systems and components needed to perform the nuclear safety functions in the event of a safe shutdown earthquake. The licensee requested approval to omit the seismic requirements of NFPA 805 Section 3.6.4 concerning the existing installation of the Class II Hose Station and Standpipe System. See SE Section 3.1.4.10 for the NRC staff's review of this request.
NFPA 805 Section 3.8.2 requires that if automatic fire detection is required to meet the performance or deterministic requirements of Chapter 4, then. these devices shall be installed in accordance with NFPA 72, National Fire Alarm Code (Reference 64), and its applicable appendixes. The licensee requested that the existing layout and placement of fire detection devices installed in accordance with the code of record (NFPA 72E-1978) remain in place even though fire detection panels were upgraded to NFPA 72 without including the relocation or re-design of detection devices to later versions of NFPA 72. See SE Section 3.1.4.11 for the NRC staff's review of this request.
In a letter dated October 14, 2013 (Reference 11 ), the licensee submitted a revised LAR Attachment L which provided additional details on the previously submitted requests for approval and also added two additional approval requests. The NFPA 805 sections identified in the revised LAR Attachment L as complying via this method are:
3.3.12(a) 3.2.3(1)
NFPA 805 Section 3.3.12(a) requires the oil collection system for reactor coolant pumps be designed and installed such that leakage from the oil system is safely contained for off normal conditions such as accident conditions or earthquakes.
The licensee requested approval for the potential of oil misting from the reactor coolant pumps due to normal motor consumption. See SE Section 3.1.4.12 for the NRC staff's review of this request.
NFPA 805 Section 3.2.3(1) requires procedures shall be established for implementation of the fire protection program. The licensee indicated that approval is requested to use a performance-based method to establish inspection, testing, and maintenance frequencies for fire protection systems and features required by NFPA 805 in accordance with 10 CFR 50.48 (c)(2)(vii). See SE Section 3.1.4.13 for the NRC staff's review of this request.
As discussed in SE Section 3.1.4 below, the NRC staff finds the use of performance-based methods to demonstrate compliance with these fundamental fire protection program elements acceptable.
3.1.1.6 Compliance Strategy - No Review Required LAR Section 4.1.2.1 identified a category of compliance strategy as "No Review Required". The category was defined by the licensee as "the existing Chapter 3 elements are not based on the requirements and/or are not applicable to elements of the VCSNS Fire Protection Program." This compliance strategy addressed NFPA 805 sections that had no technical requirements, had technical requirements that did not apply to the licensee's FPP since the element/system/attribute is not used at VCSNS, or had options to comply that were not applicable. In these instances, with the exception of NFPA 805 Section 3.11.3(3), based on the information provided by the licensee, the NRC staff concludes that the licensee's statements of compliance are acceptable.
The licensee identified NFPA 805 Section 3.11.3(3) in the LAR B-1 Table as complying via this method as a result of its conclusion that all requirements of NFPA 101 were not applicable to RI/PB FPPs in accordance with 10 CFR 50.48(c)(i). The staff disagreed with the licensee's conclusion. NFPA 805, Section 3.11.3(3) "Fire Barrier Penetrations", requires that for fire barrier penetrations the licensee comply with NFPA 101, Life Safety Code (Reference 67). In FPE RAI 18 (Reference 20) the NRC staff requested that the licensee identify the compliance strategy for penetrations in fire barriers. In its response to FPE RAI 18 (Reference 8), the licensee stated that the code compliance document and Table B-1 would be revised to incorporate applicable sections of NFPA 101 related to maintaining the integrity of fire rated barriers via protection of personnel and equipment openings including fire protection devices such as doors and dampers.
The NRC staff found that LAR Attachment S, Table S-2, includes Items 11 and 12 that address updating LAR Attachment A, Table B-1 as well as surveillance procedures and installation details to bring VCSNS into compliance with NFPA 805, Section 3.11. The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee identified a required action that will incorporate the provisions of NFPA 805, Chapter 3 in the licensee's fire protection program and included the action as an implementation item in LAR Attachment S, which would be required by the proposed license condition.
3.1.1.7 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 element.
In each of these cases, based on the information provided by the licensee in the LAR as supplemented, and the information obtained during the site audit (ADAMS Accession No. ML15041A168), the NRC staff concludes that the individual compliance statements are acceptable. The NRC staff also concludes that the combination of compliance strategies is acceptable, and that holistic compliance with the fundamental fire protection program element is assured.
3.1.1.8
- Chapter 3 Sections Not Reviewed Some NFPA 805 Chapter 3 sections either do not apply to the transition to a risk-informed, performance-based fire protection program 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:
3.1.1.9
- Sections that do not contain any technical requirements. (e.g., NFPA 805, Section 3.1, 3.4.5, and 3.11 ).
Sections that are not applicable because of the following:
o The licensee stated that they do not have systems of this type installed.
(e.g., NFPA 805, Section 3.5.2 which applies to fire water tanks);
o The type of system is not required under the risk-informed, performance-based fire protection program (e.g., NFPA 805, Section 3.9.1 (2), which applies to water spray systems and Section 3.10.1 (2),
which applies to Halon based fire extinguishing systems); and o
The requirements are structured with an applicability statement (e.g.,
NFPA 805, Section 3.4.1 (a)(2) and Section 3.4.1 (a)(3), which applies to the fire brigade standards used since they depend on the type of brigade specified in the fire protection program).
Compliance with Chapter 3 Requirements Conclusion As discussed above, the NRC staff evaluated the results of the licensee's assessment of the proposed risk-informed, performance-based fire protection program against the NFPA 805, Chapter 3, fundamental fire protection program elements and minimum design requirements, as modified by the exceptions, modifications, and supplementations in 10 CFR 50.48(c)(2). Based on.this review of the licensee's submittal, as supplemented, the NRC staff concludes that the risk-informed, performance-based fire protection program is acceptable with respect to the fundamental fire protection program elements and minimum design requirements of NFPA 805, Chapter 3, as modified by 10 CFR 50.48(c)(2), because the licensee:
Used an overall process.consistent with NRC staff approved guidance to determine the state of compliance with each of the applicable NFPA 805 Chapter 3 requirements.
Provided appropriate documentation of the state of compliance with the NFPA 805 requirements, which adequately demonstrated compliance in that the licensee was able to substantiate that it complied:
o With the requirement directly or with the requirement directly after the completion of an implementation item; o
With the intent of the requirement (or element) and provided adequate justification; o
Via previous NRC staff approval of an alternative to the requirement; o
Through the use of an engineering equivalency evaluation; o
Through the use of a combination of the above methods; and o
Through the use of a performance-based method that the NRC staff has specifically approved in accordance with 10 CFR 50.48(c)(2)(vii).
3.1.2 Identification of Power Block The NRC staff reviewed the structures identified in LAR Table 1-1 "VCSNS Power Block Definition" as supplemented, as comprising 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 risk-informed, performance-based fire protection program that have additional requirements in accordance with 10 CFR 50.48(c) and NFPA 805. As stated in 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 LAR also notes that structures meeting the radioactive release criteria described in NFPA 805, Section 1.5, but not required for nuclear plant operations, are separately screened and included in the radioactive release review as discussed in LAR Section 4.4 and LAR Attachment E.
During the review, the NRC staff observed that certain structures in the YARD were not identified in LAR Attachment I, Table 1-1. In FPE RAI 16 (Reference 20) the NRC staff requested that the licensee confirm the power block included certain structures in the YARD, or provide a justification for their exclusion. In its response to FPE RAI 16 (Reference 8), the licensee confirmed that the power block structure included the following:
Condensate Storage Tank; Refueling Water Storage Tank; Manholes; Diesel Generator Fuel Oil Tanks; Auxiliary Boiler Oil Storage Tanks; and Transformers.
The NRC staff concludes that the licensee has appropriately evaluated the structures and equipment 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 MT' Fire Barrier Configurations" Issues 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. In a letter dated June 5, 2006 (Reference 100), the licensee stated that Hemyc TM or MT' electrical raceway fire barrier systems (ERFBS) are not relied upon for separation and/or safe shutdown purposes, and are not credited in any Station analyses. Since Hemyc or MT ERFBS are not used, the NRC staff concludes that the generic issue, GL 2006-03 (Reference 51A) related to the use of these ERFBS is not applicable.
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 performance-based methods permitted elsewhere in the standard as a means of demonstrating compliance with the prescriptive fire protection program fundamental elements and minimum design requirements of NFPA 805 Chapter 3. Paragraph 50.48(c)(2)(vii) of 10 CFR requires that an acceptable performance-based 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).
NFPA 805, Section 1.3.1 "Nuclear Safety Goal," states that:
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.
NFPA 805, Section 1.3.2 "Radioactive Release Goal," states that:
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.
NFPA 805, Section 1.4.1, "Nuclear Safety Objectives," states that:
In the event of a fire during any operational mode and plant configuration, the plant shall be as follows:
( 1)
Reactivity Control. Capable of rapidly achieving and maintaining subcritical conditions.
(2)
Fuel Cooling. Capable of achieving and maintaining decay heat removal and inventory control functions.
(3)
Fission Product Boundary. Capable of preventing fuel clad damage so that the primary containment boundary is not challenged.
NFPA 805, Section 1.4.2 "Radioactive Release Objective," states that:
Either of the following objectives shall be met during all operational modes and plant configurations.
( 1)
Containment integrity is capable of being maintained.
(2)
The source term is capable of being limited.
NFPA 805, Section 1.5.1 "Nuclear Safety Performance Criteria," states that:
Fire protection 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 demonstrate this, the following performance criteria shall be met.
(a)
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 PWR and shall be capable of maintaining or rapidly restoring reader water level above top of active fuel for a 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.
(e)
Process Monitoring. Process monitoring shall be capable of providing the necessary indication to assure the criteria addressed in (a) through (d) have been achieved *and are being maintained.
NFPA 805, Section 1.5.2 "Radioactive Release Performance Criteria," states that:
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 LAR Attachment L, "NFPA 805, Chapter 3, Requirements for Approval 10 CFR 50.48(c)(2)(vii)," the licensee requested NRC staff review and approval of performance based methods to demonstrate an equivalent level of fire protection for the requirement of the elements identified in SE Section 3.1.1.5. In FPE RAI 14-02 (Reference 21 ), the NRC staff requested that the licensee provide additional details to explicitly describe how safety margin and defense-in-depth are specifically maintained for each approval request. In its response to FPE RAI 14-02 (Reference 11), the licensee submitted a revised LAR Attachment L, "Request for Approval," which included two additional "Requests for Approval" not included in the original LAR Attachment L. The NRC staff evaluation of these proposed methods is provided below.
3.1.4.1 Use of Non-treated Wood in Limited Quantities In LAR Attachment L, Approval Request L 1, the licensee requested NRC staff review and approval of a performance-based method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.3.1.2(1) regarding use of non-treated wood in limited quantities. Specifically, the licensee stated that field conditions may be experienced where non~treated wood may be needed to address unique situations during plant operations or during outages.
The licensee stated that the basis for this request is to address instances where minor non-compliances of use of non-treated wood in limited quantities may be necessary. The licensee further stated that administrative procedures may permit this condition based on added compensatory measures, additional engineering approvals or other administrative actions to manage the conditions and minimize the risk.
In FPE RAI 05 (Reference 20), the NRC staff requested that the licensee provide a description of the process that would control the use of untreated wood. In its response to FPE RAI 05 (Reference 8) the licensee stated that plant maintenance and operations will at times require untreated wood in the form of small hand tools and maintenance equipment and untreated wood for temporary material/equipment transport (e.g., pallets or equipment crates) and that the transient/combustible control program will address deviations from this Chapter 3 requirement by identifying the limits and controls for these conditions when necessary. The licensee further stated that the bounding limits of these controls will control the amount, duration, and compensatory measures for the use of untreated wood and will be provided on a specific fire scenario basis in order to maintain the assumptions in the Fire PRA and that any changes to these bounding limits will be processed via the "Plant Change Evaluation" to maintain the integrity of the Fire PRA. The licensee further stated that non-compliances with these administrative controls will be addressed as program non-compliances under the plant's corrective action program with condition evaluations and corrective actions taken to address and restore compliance and that the established administrative controls will be bounded by the combustible quantities and heat release rates assumed in the Fire PRA. The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated the use of an appropriate process to control the use of untreated wood.
The licensee stated that the use of limited amounts of untreated wood.in selected risk significant areas is restricted by administrative and engineering procedures with suitable fire protection features present in the area that ensure for the control of transient combustibles, separation distance, suppression, fire barriers and protection of the nuclear safety performance criteria as applicable and identified by VCSNS and NFPA 805 Section 1.5. The licensee further stated that the use of combustible materials such as wood in a radiological area is closely reviewed and limited due to potential effects of fire and ALARA and that there is no nuclear safety or radiological concern from transient non-treated wood that is not under strict review and controls.
The licensee stated that the use of limited amounts of untreated wood in selected risk significant areas are controlled by administrative and engineering procedures. The licensee further stated that the controls in place to restrict the usage of untreated wood meet the intent of the NFPA 805 Chapter 3 requirement, (which is that transient combustible materials in the power block be controlled), and therefore, the safety margin that is inherent within the NFPA 805 Chapter 3 requirement is maintained.
The licensee stated that the three major elements of defense-in-depth are: (1) prevent fires from starting, (2) rapidly detect, control and promptly extinguish any fires that do occur, thereby limiting fire damage, and (3) provide sufficient fire protection for structures, systems and components important to safety so that, in the event a fire is not extinguished promptly, it will not prevent essential plant safety functions from being performed. The licensee further stated that the use of limited amounts of untreated wood in selected risk significant areas does not affect element (1) as there are strict administrative and engineering controls for transient combustibles within the power block. Elements (2) and (3) are not affected by the use of limited amounts of untreated wood as there would be no effect on the fire detection, automatic and manual fire suppression activities, and the ability of the barriers to restrict the passage of smoke and flame, and therefore the defense-in-depth measures are maintained.
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 performance-based method is an acceptable alternative to the corresponding NFPA 805, Section 3.3.1.2(1) 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 defense-in-depth.
3.1.4.2 Limited Unqualified Wiring Above Suspended Ceilings In LAR Attachment L, Approval Request L2, the licensee requested NRC staff review and approval of a performance-based method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.3.5.1 for existing wiring in suspended ceilings.
Specifically, the licensee stated that while the code section is prescriptive in the use and limitation of exposed electrical wire above suspended ceilings, there is existing wiring for non-essential, non-risk significant areas and systems such as lighting and electrical power outlets that may not meet the literal requirements of this section for those limited areas of the plant with suspended ceilings.
In FPE RAI 08 (Reference 20), the NRC staff requested more detail regarding the existing wiring above the suspended ceilings and the justification for hazards mitigation. In its response to FPE RAI 08 (Reference 8), the licensee stated that the predominant cable use routed above these ceilings is for lighting and that this configuration consists of conduit routed to a junction box with armored flexible cable from the junction box to the light. The licensee further stated that most other circuits such as fire/smoke detectors and ventilation are routed in conduit with no allowance of open tray installation in these overhead areas and that the only installations allowed above the suspended ceilings are enclosed tray or square conduit, with one exception of communication cables (phone, computer and security). The licensee further stated that communication cables are not routed with power/control cables and that they are rarely routed in conduit, but due to their low energy are not considered a potential source of fire. The licensee further stated that these exposed cables predomina.ntly have PVC jacketing which presents an additional smoke concern for a fire, but is not considered to be a contributing factor to potential fire spread because these cables are routed through fire seals at fire barriers similar to other cables. The licensee further stated that the majority of areas with suspended ceilings are designed with fire/smoke detectors located above and below the suspended ceiling with corresponding sprinkler systems and that most cables located above suspended ceilings meet the NFPA 805 Chapter 3 requirement but, there is a limited installation of noncompliant communication cables, whose additional hazards are mitigated by the presence of fire seals, fire/smoke detectors located either above or below the suspended ceilings and associated sprinkler systems. The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee provided appropriate justification for hazards mitigation for existing wiring above suspended ceilings.
The licensee stated that the use of limited amounts of wiring above suspended ceilings in selected risk significant areas is restricted by engineering specifications and procedures with suitable fire protection features present in the area to ensure the control of combustibles, separation distance, suppression, fire barriers and protection of the nuclear safety performance criteria as applicable and identified by VCSNS and NFPA 805 Section 1.5. The licensee further stated that the existence of wiring above suspended ceilings or in a radiological area is closely reviewed and limited due to potential effects of fire and ALARA and that there is no nuclear safety or radiological concern from wiring above suspended ceilings that is not under strict review and engineering controls.
The licensee stated that the installation of wiring above suspended ceilings is limited and that where installed, the wiring is specified to be enclosed within metal conduits, cable trays, armored cables, or plenum rated through fire testing. The licensee further stated that the controls in place to restrict the installation of wiring above suspended ceilings and to protect the wiring when installed meet the intent of the NFPA 805 Chapter 3 requirement (which is to limit the combustible materials above suspended ceilings), and therefore, the safety margin that is inherent within the NFPA 805 Chapter 3 requirement is maintained.
The licensee stated that the three major elements of defense-in-depth are: (1) prevent fires from starting, (2) rapidly detect, control and promptly extinguish any fires that do occur, thereby limiting fire damage, and (3) provide sufficient fire protection for structures, systems and components important to safety so that, in the event a fire is not extinguished promptly, it will not prevent essential plant safety functions from being performed. The licensee stated that the installation of wiring above suspended ceilings does not affect element (1) as there are limited areas which contain suspended ceilings, and there are specifications which govern the installation of wiring above suspended ceilings in these areas. The licensee further stated that Elements (2) and (3) are not affected by the installation of wiring above suspended ceilings as the limited installation and specifications for enclosure and fire testing of the wiring does not impact the functions of the fire detection, automatic and manual fire suppression activities, and the ability of the barriers to restrict the passage of smoke and flame and therefore the defense-in-depth measures are maintained.
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 performance-based method is an acceptable alternative to the corresponding NFPA 805, Section 3.3.5.1 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 defense-in-depth.
3.1.4.3 Electrical Cable Construction with N.on-compliant Flame Propagation Tests In LAR Attachment L, Approval Request L3, the licensee requested NRC staff review and approval of a performance-based method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.3.5.3 that requires electrical cable construction comply with a flame_ propagation test as acceptable to the Authority Having Jurisdiction (AHJ). In the response to FPE RAI 14.01.b.i (Reference 11), the licensee stated that cables tested by more current test methods may have similar or better flame propagation resistance than if tested by institute of Electrical and Electronics Engineers (IEEE) 383-1974 test method and that these alternative flame propagation test methods may be utilized when verifying and validating new electrical cable when purchased at VCSNS prior to field installation.
The licensee stated that IEEE 383 (Reference 71) standard was selected as the baseline since it has been previously referenced as the NRC minimum test standard and acceptance criteria for cable flame propagation tests and that the NRC provided alternative test standards as input to industry FAQ 06-0022 (Reference 72) generated by the NFPA 805 transition process.
In FPE RAI 09 (Reference 20), the NRC staff requested that the licensee provide additional detailed information regarding the as-built condition of currently installed "existing non-compliant" cables, and to also describe the cable flame spread or other cable construction standards that were used for the currently installed cables. In its response to FPE RAI 14.01.b.i (Reference 11 ),
the licensee stated that based on the known and approximated values of qualified cable installed in the VCSNS power production areas of the plant, the non-IEEE 383-1974 cable installed in the plant comprises less than 2.5% of the total linear feet of cable installed. The licensee further stated that the majority of the unqualified cable is communication cabling (phone, computer or security) and therefore is routed predominantly through/to areas that do not impact NFPA 805 support circuits/equipment since these support circuits/equipment are not located in these areas.
The licensee further stated that where communication cabling is located in a potentially impacting area, these areas are continuously manned or consist of office areas (i.e.; Control Room, Technical Support Center, Engineering, or Security). The licensee further stated that the majority of the non-IEEE 383-197 4 cabling is routed in conduit and that non-qualified cable which is routed in conduit is considered to meet the intent of IEEE 383-1974 due to the reduced availability of oxygen in conduit which will limit the potential fire growth and flame spread. In addition, the lack of openings in the conduit restricts the exit of any smoke generated. The licensee further stated that the major impact of non-IEEE 383-197 4 cables installed in the plant is the additional smoke hazard due to the polyvinyl chloride.(PVC) of the cables in a fire scenario and that the non-qualified cables in conduit do not contribute to additional smoke or fire probability. The licensee further stated that any additional smoke hazard from the PVC is mitigated by the fact that the VCSNS fire brigade dons self-contained breathing apparatuses for all firefighting efforts and that with a large portion of the non-qualified cables located in plant office areas (Operations, Engineering, or Security), the potential impact of these cables is further reduced as a fire event in these areas is likely to be disco,vered in the incipient phase. The licensee further stated that there currently are no situations that involve the use of non-qualified cables routed within train specific non-safety related cable trays or conduits. The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee provided adequate justification for the continued use of non-IEEE 383 cables.**
The licensee stated that the use of existing test methods and/or new test methods to assess the behavior of assemblies and/or materials is always developing with technology and would not present a nuclear safety or radiological concern from utilizing an alternative approach that is
- performance based. The licensee further stated that these methods are reviewed by a qualified fire protection engineer(s) that is knowledgeable with the Fire PRA methodology and the risk significant areas of the plant.
The licensee stated that the use of fire test standards for qualification of electrical cable is evaluated in FAQ 06-0022 and is approved by the NRC for use by licensees in transition. The licensee further stated that VCSNS monitors the purchase and use of electrical cable for compliance with the fire testing allowed in FAQ 06-0022 and that these controls meet the intent of the NFPA 805 Chapter 3 requirement (which is that electric cable construction be qualified in accordance with a flame propagation test acceptable to the AHJ), and therefore, the safety margin that is inherent within the NFPA 805 Chapter 3 requirement is maintained.
The licensee stated that the three major elements of defense-in-depth are: (1) prevent fires from starting, (2) rapidly detect, control and promptly extinguish any 'fires that do occur, thereby limiting fire damage, and (3) provide sufficient fire protection for structures, systems and components important to safety so that, in the event a fire is not extinguished promptly, it will not prevent essential plant safety functions from being performed. The licensee stated that the use of fire test standards for qualification of electrical cable evaluated in FAQ 06-0022 does not affect elements (1 ), (2) or (3) and that the usage of these fire tests to qualify electrical cables used at VCSNS does not impact VCSNS's ability to prevent fires, the functions of the fire detection, automatic and manual fire suppression activities, or the ability of the barriers to restrict the passage of smoke and flame and therefore, the defense-in-depth measures are maintained.
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 performance based method is an acceptable alternative to the corresponding NFPA 805, Section 3.3.5.3 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 fire protection defense-in-depth.
3.1.4.4 Bulk Gas Storage Tanks In LAR Attachment L, Approval Request L4, the licensee requested NRC staff review and approval of a performance-based method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.3.7.2 that requires that outdoor high-pressure flammable gas storage containers be located so that the long axis is not pointed at buildings. Specifically, the licensee requested approval for the existing horizontal hydrogen storage tanks that are perpendicular to the Turbine Building/Control Building wall. The licensee further stated that the request is based on approximately 240 feet of separation distance and that the substantial
v distance of the hydrogen storage tanks from the Turbine and Control buildings is an alternative approach to the prescriptive requirement of the code regarding the orientation of the tank axis.
In FPE RAI 10 (Reference 20), the NRC staff requested more detailed information regarding the extent of the hazard, the quantity and capacity of the tanks, a description of the refilling process and a description of the features that reduce the hazards associated with the alternative method.
In its response to FPE RAI 10 (Reference 8), the licensee stated that the bulk hydrogen storage facility consists of six hydrogen storage cylinders located outside the protected area south of the turbine building and that the cylinders are maintained at 2400 psig, and the tanks are filled at the storage location by a vendor using a tube trailer in accordance with site procedures. The licensee further stated that the tanks are surrounded by fencing and cement vehicle barrier posts and that there are no missile shields surrounding the tanks. The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee provided adequate justification for the alternative method.
The licensee stated that an explosion of one of the bulk hydrogen storage tanks may result in a release of.energy equivalent to 348 lb-TNT, that a 66 ounce fragment missile traveling at 1965 ft/sec could be generated by such an explosion, and that such a missile could penetrate up to 8 inches into a concrete structure.* The licensee further stated that all Seismic Category I structures at this site are 2 feet thick, double re-enforced; therefore, this missile may result in spalling and cracking of the concrete, but will not penetrate and therefore not damage components within these structures.
The licensee stated that these tanks are located in the exterior yard and there is no radiological or nuclear safety concern.
The licensee stated that, the existing horizontal, hydrogen storage tanks are oriented perpendicular to the Turbine Building/Control Building and are located approximately 240 feet away from the Turbine Building and another 100 feet to the Control Building. The licensee further stated that based upon a review of NFPA codes and standards as well as NEIL requirements, the greatest required distance for location of outdoor hydrogen storage tanks is 50 feet away from potential exposures and that the location of the tan.ks approximately 240 feet away provides a significant safety margin (approximately 190 feet to the Turbine Building). The licensee further stated that the tanks are bolted in place to limit potential movement, that the location and securing of the hydrogen storage tanks meets the intent of the NFPA 805 Chapter 3 requirement (which is to minimize the damage to buildings should a storage container rupture), and therefore, the safety margin that is inherent within the NFPA 805 Chapter 3,requirement is maintained.
The licensee stated that the three major elements of defense-in-depth are: ( 1) prevent fires from starting, (2) rapidly detect, control and promptly extinguish any fires that do occur, thereby limiting fire damage, and (3) provide sufficient fire protection for structures, systems and components important to safety so that, in the event a fire is not extinguished promptly, it will not prevent essential plant safety functions from being performed. The licensee further stated that the location of the existing horizontal hydrogen storage tanks that are oriented perpendicular to the Turbine Building/Control Building does not affect elements (1 ), (2) or (3) and that there is no impact to VCSNS's ability to prevent fires and the functions of the fire detection and automatic and manual fire suppression activities. The licensee further stated that the location of the hydrogen storage tanks approximately 240 feet from the nearest required barrier maintains the ability of the barriers to restrict the passage of smoke and flame and therefore, the defense-in-depth measures are maintained.
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 performance based method is an acceptable alternative to the corresponding NFPA 805, Section 3.3.7.2 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 defense-in-depth.
3.1.4.5 Fire Brigade Notification In LAR Attachment L, Approval Request LS, the licensee requested NRC staff review and approval of a performance-based method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.4.1 (d) that requires the fire brigade be notified immediately upon verification of a fire. The licensee indicated that the control room notifies the fire brigade upon verification of a fire and that verification could be accomplished by several methods and at VCSNS verification is made by direct visual contact with the fire and/or products of combustion and with direct communication to the control room.
The licensee stated that its approach allows for the immediate dispatch of someone from operations to the scene of the alarm signal, perform verification and begin to assess the status and potential effects to nuclear safety. The licensee further stated that action is the verbal confirmation back to the control room that dispatches the fire brigade and brigade leader with knowledge of its specific location and its potential which allows brigade members and.the control room immediate and credible information to act without delay to alleviate smoke and heat conditions, protect equipment and advance hose lines, as necessary.
In FPE RAI 11 (Reference 20), the NRC staff requested that the licensee provide additional detail to identify the sequence of notification factored into the assumptions for time-to-damage and non-suppression probability for the Fire PRA assumptions. In its response to FPE RAI 11 (Reference 8), the licensee stated that the time to "direct visual contact with the fire" is not accounted for as an explicit parameter in the Fire PRA, that the analysis uses the methods of NUREG/CR-6850 Supplement 1 (Reference 43) for the manual non-suppression probability, and that the fire brigade response time does not affect evaluation. The licensee further stated that under this approach, the "direct visual contact with the fire" is_an action after "fire detection" and actions to control and/or suppress the fire start at detection. The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee provided adequate justification for not accounting for "the time to direct visual contact with the fire" in the fire PRA and because the licensee's use of manual suppression probability is consistent with NRC endorsed guidance.
The licensee stated that the sequence of notification that is performed allows for expedited strategic response to the fire scene and would not impact nuclear safety or create a radiological concern.
The licensee stated that the notification of the fire brigade upon visual confirmation of a fire event provides information speCific to the event and that this sequence allows the fire brigade to respond appropriately to the fire event and alleviates the potential for unnecessary actions. The
. licensee further stated that since the fire brigade consists of a minimum of five (5) individuals, notification to the Control Room allows the Control Room to use the plant public address system to communicate with all fire brigade members at once and that this process ensures that all fire brigade members are alerted simultaneously regardless of their individual location within the plant site. The licensee further stated that the fire brigade procedures and training ensure the prompt availability and actions of the fire brigade in a fire event and that the visual verification of a fire and subsequent verbal notification of the fire brigade meets the intent of the NFPA 805 Chapter 3 requirement (which is to ensure on site firefighting capability), and therefore, the safety margin that is inherent within the NFPA 805 Chapter 3 requirement is maintained.
The licensee stated that the three major elements of defense-in-depth are: (1) prevent fires from starting, (2) rapidly detect, control and promptly extinguish any fires that do occur, thereby limiting fire damage, and (3) provide sufficient fire protection for structures, systems and components important to safety so that, in the event a fire is not extinguished promptly, it will not prevent essential plant safety functions from being performed. The licensee stated that the notification of the fire brigade upon visual confirmation of a fire event does not affect elements (1 ), (2) or (3) and that there is no impact on VCSNS's ability to prevent fires or the functions of the fire detection and automatic fire suppression activities. The licensee further stated that the prompt notification of the fire brigade and appropriate manual actions to contain the fire and maintain the ability of the barriers to restrict the passage of smoke and flame remain unaffected, and therefore, the defense-in-depth measures are maintained.
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 performance-based method is an acceptable alternative to the corresponding NFPA 805, Section 3.4.1 (d) 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 defense-in-depth.
3.1.4.6 Pre-Fire Plans In LAR Attachment L, Approval Request L6, the licensee requested NRC staff review and approval of a performance-based method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.4.2.4 that requires that Pre-fire plans address coordination with other plant groups during fire emergencies. The licensee requested approval for the use of multiple procedures to coordinate the fire brigade activities with other groups. The licensee stated that the pre-fire plan, emergency procedures and brigade leader training assures the required coordination and that the use of pre-fire plans considers the coordination of support groups and training is provided on many scenarios that would include a variety of other groups.
The licensee further stated that in some instances in drills and/or in an ongoing event the need to interact with specific groups would be driven by variables that may not be predictable.
The licensee stated that the station Emergency Plan (EP) procedures and fire brigade Leader Training discuss coordination with other groups during fire emergencies and that the coordination
- with support groups may not be located within the context of nor need to be located within the "Pre Fire Plans", since the EP procedures address interfaces and support.
The licensee stated that the procedural location of specific coordination of a fire support group(s) would not impact nuclear safety or create a radiological concern from utilizing an alternate approach that is effective and performance based.
The licensee stated that the coordination of the fire brigade with other plant groups is achieved through the use of the pre-fire plans and emergency procedures and that given the unique nature of a fire event, the active coordination process between the fire brigade and emergency plan coordinators allows for a dynamic response to any fire event. The licensee further stated that the fire brigade procedures, emergency plan procedures, pre-fire plans, and training ensure that coordination between the fire brigade and other groups is accomplished in a fire event and that this process meets the intent of the NFPA 805 Chapter 3 requirement and therefore, the safety margin that is inherent within the NFPA 805 Chapter 3 requirement is maintained.
The licensee stated that the three major elements of defense-in-depth are: ( 1) prevent fires from starting, (2) rapidly detect, control and promptly extinguish any fires that do occur, thereby limiting fire damage, and (3) provide sufficient fire protection for structures, systems and components important to safety so that, in the event a fire is not extinguished promptly, it will not prevent essential plant safety functions from being performed. The licensee stated that the coordinati9n of the fire brigade with other plant groups through the use of the pre-fire plans and emergency procedures does not affect elements (1 ), (2) or (3) and that there is no impact on VCSNS's ability to prevent fires or the functions of the fire detection and automatic fire suppression activities. The licensee further stated that the prompt notification and actions of the fire brigade and the ability of the barriers to restrict the passage of smoke and flame are maintained and therefore, the defense-in-depth measures are maintained.
In accordance with 10 CFR 50.48(c)(2)(vii}, the NRC staff finds the proposed performance-based method is acceptable for application in lieu of the corresponding NFPA 805, Section 3.4.2.4 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 margins, and maintains adequate fire protection defense-in-depth.
3.1.4.7 Records/
In LAR Attachment L, Approval Request L7, the licensee requested NRC staff review and
- approval of a performance-based method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.4.3(a)(4) that requires written records documenting fire brigade drills and training. The licensee requested the use of electronic records and or written records that document fire brigade member training in lieu of "written records". The licensee stated that the primary storage medium for these training records is electronic, and "written records" are typically not maintained and that the subject training records may be paperless media that is available and controlled by the station's Record Management System.
The licensee stated that the storage medium of records would not impact a nuclear safety or create a radiological concern from utilizing an alternate approach that is effective and performance based.
The licensee stated that the documentation of fire brigade member training is maintained through electronic media in lieu of written records and that procedures are in place to control the recording of fire brigade member training as well as securing and storing the information in the station's Record Management System. The licensee further stated that the use of electronic records instead of written records meets the intent of the NFPA 805 Chapter 3 requirement (which is to ensure adequate fire brigade training and drills), and therefore, the safety margin that is inherent within the NFPA 805 Chapter 3 requirement is maintained.
The licensee stated that the three major elements of defense-in-depth are: (1) prevent fires from starting, (2) rapidly detect, control and promptly extinguish any fires that do occur, thereby limiting fire damage, and (3) provide sufficient fire protection for structures, systems and components important to safety so that, in the event a fire is not extinguished promptly, it will not prevent essential plant safety functions from being performed. The licensee stated that documentation of fire *brigade member training through electronic media in lieu of written records does not affect elements (1 ), (2) or (3) and that there is no impact on VCSNS's ability to prevent fires, the functions of the fire detection, automatic and manual fire suppression activities, or the ability of the barriers to restrict the passage of smoke and flame and therefore, the defense-in-depth measures are maintained.
In accordance with 10 CFR 50.48(c)(2)(vii), the NRC staff finds the proposed performance-based method is acceptable for application in lieu of the corresponding NFPA 805, Section 3.4.3(a)(4) 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 margins, and. maintains adequate fire protection defense-in-depth.
3.1.4.8 Yard Fire Hydrant Layout In LAR Attachment L, Approval Request LB, the licensee requested NRC staff review and approval of a performance-based method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.5.15 that requires that hydrants be installed approximately every 250 feet apart on the yard main system. The licensee requested a performance-based hydrant separation scheme based on an average of approximately 325 feet of separation between hydrants protecting buildings and structures within the protected area.
The licensee stated that the intent of this requirement, as specified in NFPA 24-1973 (Reference 73), is to locate fire hydrants such that a sufficient number of hydrants are provided for exterior and interior firefighting and that NFPA 24 indicates two hose streams for every part of the interior of each building not covered by standpipe protection and a single hose stream to protect the exterior of buildings with interior standpipe systems. The licensee stated that both requirements specify that there shall be sufficient hydrants to concentrate the required fire flow about any important building with no hose line exceeding 500 feet in length. The licensee stated that Appendix A to Branch Technical Position 9-5.1 (Reference 7 4) indicates that "Outside manual hose installation should be sufficient to reach any location with an effective hose stream.
To accomplish this, hydrants should be installed approximately every_ 250 feet on the yard main system". Th.e licensee further stated that this approximate distance is recommended, but may not be necessary, in order to accomplish this intent of this requirement.
The licensee stated that a review of plant drawings and plant walkdowns has confirmed that there
. is a s~fficient number of yard fire hydrants located such that two hose streams with hose lengths of 500 feet or less (from single or multiple hydrants) can reach the interior buildings not provided with interior standpipe systems. The licensee further stated that the remaining buildings are provided with a sufficient number of Class II standpipes located throughout the structure to enable the fire brigade to reach all areas of the plant by an interior hose stream and that the review of plant drawings and plant walkdowns has also confirmed that there is a sufficient number of yard fire hydrants located such that a hose stream with hose lengths of 500 feet or less can reach the exterior of each of these buildings.
The licensee stated that the current spacing of yard fire hydrants meets the intent of NFPA 24-1973 and is considered to provide a functional equivalency to the approximate spacing specified in the codes. The licensee further stated that the current layout of yard fire hydrants would not impact nuclear safety and that the fire hydrants are located on the yard main and would not impact radiological release performance criteria.
The li<:;ensee stated that the fire hydrants are located with an approximate average spacing of 325 feet and that evaluations for the fire hydrants indicate sufficient coverage of the interior of buildings not equipped with Class II hose stations as well as the exterior of buildings. The licensee further stated that the approximate average hydrant spacing of 325 feet combined with the 500 feet of hose line provides a functional equivalency to the NFPA 805 Chapter 3 requirement and therefore, the safety margin that is inherent within the NFPA 805 Chapter 3 requirement is maintained.
The licensee stated that the three major elements of defense-in-depth are: (1) prevent fires from starting, (2) rapidly detect, control and promptly extinguish any fires that do occur, thereby limiting fire damage, and (3) provide sufficient fire protection for structures, systems and components important to safety so that, in the event a fire is not extinguished promptly, it will not prevent essential plant safety functions from being performed. The licensee stated that the location of fire hydrants with an approximate average spacing of 325 feet does not affect elements (1 ), (2) or (3) and that there is no impact on VCSNS's ability to prevent fires or the functions of the fire detection and automatic fire suppression activities. The licensee stated that the coverage provided by the fire hydrants and hose lines is adequate to support the fire brigades' manual suppression actions and the ability of fire barriers to restrict the passage of smoke and flame is not impacted and therefore, the defense-in-depth measures are maintained.
Based on its review of the information submitted by the licensee, and in accordance with 1 O CFR 50.48( c)(2)(vii), the NRC staff concludes that the proposed performance-based method is an acceptable alternative to the corresponding NFPA 805, Section 3.5.15 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 defense-in-depth.
3.1.4.9 Hose Stations - Pressure Reducers In LAR Attachment L, Approval Request L9, the licensee requested NRC staff review and approval of a performance-based method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.6.2 that requires that capability be provided to ensure adequate water flow rate and nozzle pressure for all hose stations. The licensee requested approval for not utilizing pressure reducers on the existing standpipe systems. The licensee stated that this is based on system calculations and the proper hose line training, fire brigade member capabilities, and off-site fire department member training with. hoses under high pressure conditions.
The licensee indicated that training on high pressure lines, including training with higher than normal pressure, addresses safety considerations contained in this section of NFPA 805. The licensee further stated that higher pressures at hose stations and at standpipe or hydrant connections support addressing B.5.b mitigation scenarios, as required by 10 CFR 50.54(hh),
and adequate flow and pressure for these hose stations and exterior hose houses.
In FPE RAI 12 (Reference 20), the NRC staff requested that the licensee provide a description of the system including the minimum and maximum calculated pressures and flows, and the impact to personnel and equipment of not having the pressure reducing valves installed. In its response to FPE RAI 12 (Reference 8), the licensee stated that each fire pump has a rated capacity to provide at least 2500 gallons per minute (gpm) at 125 psig and that the maximum pressure that can be produced is 150 psig, which is the fire pump relief valve setpoint. The licensee further stated that the hydrants used for fire brigade training, both initial and refresher, at the South Carolina Fire Academy are supplied by two fire water pumps rated to provide 1500 gpm each, and are computer programmed to supply a constant water pressure of 150 psi and that fire brigade members are trained and qualified on the use of the higher pressure (> 125 psig) hose lines. The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee provided adequate justification for not using pressure reducing valves.
The licensee stated that the lack of pressure reducing devices would not impact the function of the hose stations, which are effective in delivering required water supply for fire-fighting, and therefore there would be no impact to nuclear safety and no radiological concern.
The licensee stated that fire brigade training procedures and off-site fire department training ensures adequate training for fire brigade members on the safe and effective use of high pressure hose lines and that site personnel who are not trained as fire brigade members do not utilize fire protection equipment. The licensee further stated that training procedures for high pressure hose lines administered to the station's fire brigade and off-site fire departments meets the safety intent of the NFPA 805 Chapter 3 requirement (which is to ensure standpipe and hose stations are appropriately installed), and therefore, the safety margin that is inherent within the NFPA 805 Chapter 3 requirement is maintained.
The licensee stated that the three major elements of defense-in-depth are: (1) prevent fires from starting, (2) rapidly detect, control and promptly extinguish any fires that do occur, thereby limiting fire damage, and (3) provide sufficient fire protection for structures, systems and components important to safety so that, in the event a fire is not extinguished promptly, it will not prevent essential plant safety functions from being performed. The licensee stated that the hose stations not equipped with pressure reducers do not affect elements (1 ), (2) or (3) and that there is no impact on VCSNS's ability to prevent fires or the functions of the fire detection and automatic fire suppression activities. The licensee further stated that adequate training for fire brigade members using the high pressure hose lines maintains the capability of the fire brigade to manually suppress a fire and that the ability of fire barriers to restrict the passage of smoke and flame also remain unaffected, and therefore, the defense-in-depth measures are maintained.
Based on its review of the information submitted by the licensee, and in accordance with 1 O CFR 50.48(c)(2)(vii), the NRC staff concludes that the proposed performance-based method is an acceptable alternative to the corresponding NFPA 805, Section 3.6.2 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 defense-in-depth.
3.1.4.10 Seismic Analyzed Hose Stations In LAR Attachment L, Approval Request L 10, the licensee requested NRC staff review and approval of a performance-based method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.6.4 that requires that provisions be made to supply water at least to standpipes and hose stations for manual fire suppression in all areas containing systems and components needed to perform the nuclear safety functions in the event of a safe shutdown earthquake (SSE). The licensee requested approval of the existing installation of the Class II Hose Station and Standpipe System that is not seismically designed.
The licensee stated that the standpipe and hose stations were designed as Class II systems utilizing 1-1/2-inch hose connections in accordance with NFPA14-1974 (Reference 75), which provides the requirements for the design attributes of the varied classes of standpipe systems.
The licensee further stated that the selection of the Class II standpipe design was based on good engineering practices and insurance guidelines in effect at the time of design and installation. In LAR Attachment A, Table B-1, Section 3.6.4, the licemsee further stated that the standpipe and water distribution system was designed consistent with the NRC Branch Technical Position APCSB 9.5-1 Appendix A which did not contain the seismic design requirement.
The existing Class II system design provides manual fire suppression to safety related and important to safety areas within the plant, and has been designed to deliver the flow and pressure requirements of NFPA 14-1974. The licensee further stated that based upon an evaluation of the building construction, occupancy, and automatic and manual fire protection features for prompt detection and suppression, the system design provides adequate manual fire suppression capabilities.
The licensee further stated that the station has preplanned alternate provisions and strategies for the loss of fire suppression in accordance with Operating License Condition 2.C(34), Mitigation Strategy License Condition and that following an SSE, these measures and guidelines may be implemented as necessary. The licensee further stated that use of non-seismic Class II hose stations and the procedures and training for restoration of fire suppression following an SSE meet the intent of the NFPA 805 Chapter 3 requirement (which is to ensure an adequate water supply to standpipe and hose stations in the event of an SSE), and therefore, the safety margin that is inherent within the NFPA 805 Chapter 3 requirement is maintained.
The licensee also stated that the use of Class II hose stations, which are not seismically designed, would not impact nuclear safety and that the utilization of the hose stations and preplanned alternate provisions and strategies for the loss of fire suppression following a SSE would not impact radiological release criteria.
The licensee stated that the three major elements of defense-in-depth are: (1) prevent fires from starting, (2) rapidly detect, control and promptly extinguish any fires that do occur, thereby limiting fire damage, and (3) provide sufficient fire protection for structures, systems and components important to safety so that, in the event a fire is not extinguished promptly, it will not prevent essential plant safety functions from being performed. The licensee stated that the Class II hose stations not seismically designed do not affect elements (1 ), (2) or (3) and that there is no impact on VCSNS's ability to prevent fires or the functions of the fire detection and automatic fire suppression activities. The licensee further stated that the Class II hose stations and station*
procedures and training for restoration of fire suppression capabilities maintain the fire brigades' ability to suppress a fire through manual suppression actions and that the ability of the barriers to restrict the passage of smoke and flame also remain unaffected and therefore, the defense-in-depth measures are maintained.
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 performance-based method is an acceptable alternative to the corresponding NFPA 805, Section, Section 3.6.4 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 margins, and maintains adequate fire protection defense-in-depth.
3.1.4.11 Fire Detection Code of Record In LAR Attachment L, Approval Request L 11, the licensee requested NRC staff review and approval of a performance-based method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.. 8.2 that requires if automatic fire detection is required to meet the performance or deterministic requirements of Chapter 4, then these devi.ces shall be installed in accordance with NFPA 72, National Fire Alarm Code (Reference 64), and its applicable appendixes. The licensee requested that the existing layout and placement of fire detection devices installed in accordance with the code of record (NFPA 72E-1978)
(Reference 76) remain in place even though fire detection panels were upgraded to NFPA 72 without including the relocation or re-design of detection devices to later versions of NFPA 72.
The licensee stated that the automatic fire detection meets the performance requirements of the listed devices installed in accordance with NFPA 72, and its applicable appendixes except for the detector spacing which is in accordance with the NFPA 72E-1978, which is the code of record and an equivalent approach.
In FPE RAI 13 (Reference 20), the NRC staff requested clarification of the intention and scope of this request. In its response to FPE RAI 13 (Reference 8) the licensee stated "the intention of this request is not to request approval for the individual locations of detectors throughout the plant per NFPA 72E. The intent of this request was to clarify that the code of record at the time of the installation of the automatic fire detectors was NFPA 72E-1978, Standard on Automatic Fire Detectors." The licensee further stated that the fire annunciation and proprietary alarm system portion of the system was upgraded in 1992 and installed in accordance with NFPA 72 - 1990 (Reference 77), the code of record at the time of the design and installation of the upgrade and that at the time of this upgrade, NFPA 72E - 1978 was considered the applicable code of record for the fire detector locations. The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee clarified the intent and scope of the request which enabled the NRC staff to complete its review.
The licensee stated that the performance of the detection devices being located per this code of record and not an alternate code would not impact a nuclear safety or create a radiological concern.
The licensee stated that the fire detection layout was conducted in accordance with the Code of Record for the station, NFPA 72E-1978 and that subsequent upgrades to the fire detection system have been completed in accordance with NFPA 72. The licensee further stated that the fire detection layout is controlled through the use of a design calculation and any revisions are required to be reviewed and approved in accordance with the station design review process. The licensee further stated that the use of NFPA 72E-197_8 forthe fire detection layout and control of revisions to the layout meet the intent of the NFPA 805 Chapter 3 requirement (which is to ensure that automatic fire detection devices be installed in accordance with NFPA 72), and therefore, the safety margin that is inherent within the NFPA 805 Chapter 3 requirement is maintained.
The licensee stated that the three major elements of defense-in-depth are: (1) prevent fires from starting, (2) rapidly detect, control and promptly extinguish any fires that do occur, thereby limiting fire damage, and (3) provide sufficient fire protection for structures, systems and components important to safety so that, in the event a fire is not extinguished promptly, it will not prevent essential plant safety functions from being performed. The licensee stated that the design of the fire detection layout in accordance with NFPA 72E-1978 does not affect element (1) and that there is no impact on VCSNS's ability to prevent fires. The licensee further stated that Elements (2) and (3) are not affected as the design of the fire detection layout is in accordance with an approved NFPA standard and that there is no impact on the functions of the fire detection, automatic and manual fire suppression activities, and the ability of fire barriers to restrict the passage of smoke and flame and therefore, the defense-in-depth measures are maintained.
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 performance-based method is an acceptable alternative to the corresponding NFPA 805, Section, Section 3.8.2 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 margins, and maintains adequate fire protection defense-in-depth.
3.1.4.12 Reactor Coolant Pumps In LAR Attachment L, Approval Request L 12, the licensee requested NRC staff review and approval of a performance-based method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.3.12 that requires the oil collection system for reactor coolant pumps (RCPs) be designed and installed such that leakage from the oil system is safely contained for off normal conditions such as accident conditions or earthquakes. The licensee requested approval for the potential of oil misting from the reactor coolant pumps due to normal motor consumption. The licensee stated that the oil collection system was designed for pressurized and non-pressurized leakage and spillage.
The licensee stated that the VCSNS oil collection system is designed and was reviewed in accordance with 10 CFR 50, Appendix R, Section 111.0 to collect leakage from credible pressurized and non-pressurized leakage sites in the reactor coolant pump oil system. The licensee further stated that this analysis was not required to assess the potential collection of oil mist as the result of normal pump/motor operation and that oil misting is not leakage due to equipment failure, but inherent in the operation of large open motors. The licensee further stated that it is normal for large motors to lose some oil through seals, and for the oil to potentially become 'atomized' by ventilation air flow and that this atomized oil mist can then collect on surfaces in the vicinity of the reactor coolant pump as the pump design is not completely sealed so as to permit airflow for cooling. The licensee further stated that the oil mist resulting from normal operation will not adversely impact the ability of the plant to achieve and maintain safe shutdown even if ignition occurred and that each primary coolant loop has a single reactor coolant pump and they are not required to achieve and maintain safe shutdown.
The licensee provided the following summary of its justification for the request:
The oil collection system is designed to collect leakage from credible pressurized and non-pressurized leakage sites in the reactor coolant pump oil system.
Oil misted from normal operation is not leakage; it is normal motor oil consumption.
Oil misted from normal operation does not significantly reduce the oil inventory.
The oil historically released as misting does not account for an appreciable heat release rate or accumulation near potential ignition sources, non-insulated reactor coolant piping, or instrumentation and valves for the Reactor Coolant System.
The reactor coolant pumps use an oil of a high flash point, over 400 degrees Fahrenheit.
Each primary coolant loop has a single reactor coolant pump, and they are not essential to achieve and maintain safe shutdown.
Continuous monitoring of oil levels for each reactor coolant pump provides prompt notification of potential leaks.
The licensee stated that the oil mist resultant from normal operations will not adversely impact nuclear safety and that the reactor coolant pumps are not required to achieve or maintain post-fire safe shutdown and therefore, there is no impact on the nuclear safety performance criteria. The licensee further stated that the potential for oil mist from the reactor coolant pumps has no impact on the radiological release perfor:mance criteria and that the radiological release review was performed based on the manual fire suppression activities in areas containing or potentially containing radioactive materials. The licensee further stated that the entire Reactor Building in which the reactor coolant pumps are located is an environmentally sealed radiological area and that the oil mist does* not add additional radiological materials to the area or challenge -system boundaries.
The licensee stated that the oil mist resultant from normal operation will not adversely impact the ability of the plant to achieve and maintain post-fire safe shutdown even if ignition occurred and that the reactor coolant pumps are not required to achieve and maintain fire safe shutdown. The licensee further stated that the existing oil collection system is designed to collect oil leakage from credible pressurized and non-pressurized leakage sites in the reactor coolant pump oil system, and the Reactor Building has been analyzed in this current configuration. The licensee further stated that the oil collection system meets the intent of the NFPA 805 Chapter 3 requirement (which is to ensure that leakage from reactor coolant pumps be controlled), and therefore, the safety margin that is inherent within the NFPA 805 Chapter 3 requirement is maintained.
The licensee stated that the three major elements of defense-in-depth are: (1) prevent fires from starting, (2) rapidly detect, control and promptly extinguish any fires that do occur, thereby limiting fire damage, and (3) provide sufficient fire protection for structures, systems and components important to safety so that, in the event a fire is not extinguished promptly, it will not prevent essential plant safety functions from being performed. The licensee stated that the potential for oil mist from normal operation of the reactor coolant pumps does not impact fire protection defense-in-depth and will not adversely impact the ability of the plant to achieve and maintain post-fire safe shutdown even if a fire occurred. The licensee stated that Element (1) is maintained by the oil collection system and reactor coolant pump design and that Elements (2) and (3) are not affected by the oil mist from normal operation as there would be no effect on the fire detection, automatic and manual fire suppression activities, and the ability of fire barriers to restrict the passage of smoke and flame and therefore, the defense-in-depth measures are maintained.
Based on its review of the information submitted by the licensee, and in accordance with 1 O CFR 50.48(c)(2)(vii), the NRC staff concludes that the proposed performance-based method is an acceptable alternative to the corresponding NFPA 805, Section, Section 3.3.12(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 margins, and maintains adequate fire protection defense-in-depth.
3.1.4.13 Procedures In LAR Attachment L, Approval Request L 13, the licensee requested NRC staff review and approval of a performance-based method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.2.3(1) that requires that procedures be implemented for inspection, testing, and maintenance of fire protection systems and features credited by the fire protection program. The licensee requested to use a performance-based method to establish inspection, testing, and maintenance frequencies for fire protection systems and features required by NFPA 805. The licensee stated that performance-based inspection, testing, and maintenance frequencies will be established as described in Electric Power Research Institute (EPRI) Technical Report TR1006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features," Final Report, July 2003 (Reference 65).
The licensee stated that this request is specific to the use of EPRI Technical Report TR 1006756 to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features credited by the fire protection program and that this request does not include the use of EPRI Technical Report TR1006756 to determine or manage the scope of the inspection, testing, and maintenance activities, which are controlled as described in LAR (Reference 6), Section 4.6, by the Monitoring Program.
The licensee stated that the target inspections, tests, and maintenance will be those activities for the NFPA 805 required fire protection systems and features and that the reliability and frequency goals will be established to ensure the assumptions in the NFPA 805 engineering analysis remain valid. The licensee stated that failure criterion will be established based on the required credited functions and will ensure those functions are maintained and that the failure probability and confidence level will be determined based on EPRI Technical Report TR1006756 guidance, VCSNS performance history and PRA analysis values. The licensee further stated that data collection and analysis will also follow EPRI Technical Report TR1006756 document guidance and that the performance monitoring will be performed in conjunction with the monitoring program required by NFPA 805, Section 2.6 and will ensure site-specific operating experience is considered in the monitoring process. The licensee further stated that where elements of the program previously exist, the program will be revised accordingly.
The licensee stated the use of performance-based inspection, test, and maintenance frequencies established in accordance with EPRI Technical Report TR-1006756 would not impact nuelear.
safety or create a radiological concern and that the availability and reliability of the fire protection systems and features are maintained based upon industry data and station specific data.
The licensee stated that, the frequencies will be established in accordance with EPRI Technical Report TR-1006756, using industry and station specific data and that the condition monitoring program outlined in LAR Section 4.6 ensures the inspection, test, and maintenance surveillance program is controlled. The licensee further stated that the use of performance-based inspection, test, and maintenance frequencies and the control of the station inspection, test, and maintenance surveillance program meets the intent of the NFPA 805 Chapter 3 requirement (which is to ensure that procedures be established for implementation of the fire protection program, including inspection, testing and maintenance of fire protection systems and features),
and therefore, the safety margin that is inherent within the NFPA 805 Chapter 3 requirement is maintained.
The licensee stated that the three major elements of defense-in-depth are: (1) prevent fires from starting, (2) rapidly detect, control and promptly extinguish any fires that do occur, thereby limiting fire damage, and (3) provide sufficient fire protection for structures, systems and components important to safety so that, in the event a fire is not extinguished promptly, it will not prevent essential plant safety functions from being performed. The licensee stated that the use of performance-based inspection, test, and maintenance frequencies does not affect elements (1 ),
(2) or (3) and that the establishment of performance-based inspection, test, and maintenance frequencies ensures the continued availability and reliability of the fire protection systems and features. The licensee stated that there is no impact on VCSNS's ability to prevent fires, the functions of the fire detection and automatic and manual fire suppression activities, or the ability of the barriers to restrict the passage of smoke and flame and therefore the defense-in-depth measures are maintained.
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 performance-based 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 margins, and maintains adequate fire protection defense-in-depth and because the licensee has developed an action to revise fire protection preventative maintenance and surveillance procedures and included that action in LAR Attachment S, Table S-2, Implementation Item 2, which would be required by the proposed license condition.
3.2 Nuclear Safety Capability Assessment Methods NFPA 805 (Reference 1) is a risk-informed, performance-based standard that allows engineering analyses to be used to show that fire protection program features and systems provide sufficient capability to meet the requirements of 10 CFR 50.48(c).
NFPA 805 Section 2.4, "Engineering Analyses," states, in part, 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... The effectiveness of the fire protection features shall be evaluated 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 defined in Section [2.5] for the VCSNS area being analyzed.
NFPA 805Chapter1 defines the goals, objectives and performance criteria that the fire protection program must meet in order to be in accordance with NFPA 805.
NFPA 805 Section 1.3.1 "Nuclear Safety Goal," states that:
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.
NFPA 805 Section 1.4.1 "Nuclear Safety Objectives, states that:
In the event of a fire during any operational mode and plant configuration, the plant shall be as follows:
(1)
Reactivity Control. Capable of rapidly achieving and maintaining subcritical conditions.
(2)
Fuel Cooling. Capable of achieving and maintaining decay heat removal and inventory control functions.
(3)
Fission Product Boundary. Capable of preventing fuel clad damage so that the primary containment boundary is not challenged.
NFPA 805 Section 1.5.1 "Nuclear Safety Performance Criteria," states that:
Fire protection 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 demonstrate this, the following performance criteria shall be met.
(a)
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.
(e)
Process Monitoring. Process monitoring shall be capable of 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 805 Nuclear Safety Capability Assessment Methods NFPA 805, Section 2.4.2, "Nuclear Safety Capability Assessment," states the following:
The purpose of this section is to define the methodology for performing a nuclear safety capability assessment. 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 nuclear safety equipment and cables.
(4)
Assessment of the ability to achieve the nuclear safety performance criteria given a fire in each fire area.
This SE section evaluates the first three of the topics listed above. SE Section 3.5 addresses the
- assessment of the fourth step.
Regulatory Guide 1.205, Revision 1 (Reference 2) endorses NEI 04-02, Revision 2 (Reference 5), and Chapter 3 of NEI 00-01, Revision 2, "Guidance for Post-Fire Safe Shutdown Circuit Analysis" (Reference 33), and promulgates the method outlined in NEI 04-02 for conducting a nuclear safety capability assessment. This NRG-endorsed method documents in a table format (i.e., NEI 04-02 Table B-2, "NFPA 805 Chapter 2 - Nuclear Safety Transition -
Methodology Review") the licensee's comparison of its post-fire safe shutdown analyses to the guidance in NEI 00-01 Chapter 3, which has been determined to address the related requirements of NFPA 805, Section 2.4.2. The NRC staff reviewed LAR Section 4.2.1, "Nuclear Safety Capability Assessment Methodology," and LAR Attachment B, "NEI 04-02 Table B Nuclear Safety Capability Assessment-Methodology Review," against these guidelines.
The licensee developed the LAR based on the guidance provided in the three guidance documents cited above. Based on the information provided in the licensee's submittal, as supplemented, South Carolina Electric & Gas Company (SCE&G) used a systematic process to evaluate the VCSNS post-fire safe shutdown analysis against the requirements of NFPA 805, Section 2.4.2, Subsections (1 ), (2), and (3), which meets the methodology outlined in the latest NRG-endorsed industry guidance.
FAQ 07-0039 (Reference 54), provides one acceptable method for documenting the comparison of the post-fire safe shutdown analysis against the NFPA 805 requirements. This method first maps the existing post-fire safe shutdown analysis to the NEI 00-01, Rev. 2, 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 post-fire safe shutdown analysis against the NFPA 805 nuclear safety capability assessment requirements using the NRG-endorsed process in Chapter 3 of NEI 00-01, Revision 2, and documenting the results of the review in LAR Attachment B, Table B-2, "Nuclear Safety Capability Assessment - Methodology Review," in accordance with the guidance of NEI 04-02, Revision 2.
The category used by VCSNS to describe alignment with the NEI 00-01, Chapter 3, attributes is as follows:
(1)
The SSA directly aligns with the attribute: noted in the LAR Table B-2 as "Aligns."
(see discussion in SE Section 3.2.1.1)
Some attributes may not be applicable to the safe shutdown analysis (for example, the attribute may be applicable only to BWRs). These are noted in the B-2 Table as "N/A."
The licensee performed the review of the NSCA to the guidance of NEI 00-01, Revision 1 (Reference 55) instead of Revision 2 as endorsed by Regulatory Guide 1.205, Revision 1. In safe shutdown analysis (SSD) RAI 01 (Reference 20), the NRC staff requested that the licensee provide a gap analysis on the differences between the alignments using NEI 00-01, Revision 1, as the basis for transitioning the NFPA 805 nuclear safety capability as indicated in NEI 04-02, versus using NEI 00-01, Revision 2, which is the current version cited in Regulatory Guide 1.205, Revision 1. In its response to SSD RAI 01 (Reference 8), the licensee stated that a gap analysis for NSCA methodology has been performed and that a site-specific calculation is being revised to incorporate the gap analysis. The NRC staff considers this action acceptable because it is included in LAR Attachment S, Table S-2, Implementation Item 14, which would be required by the proposed license condition. The licensee further stated that it was determined that no
. adverse or non-conservative conditions were identified with respect to revised methods in Revision 2 of NEI 00-01, as applicable to NFPA 805. The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated that there were no adverse or non-conservative conditions resulting from the use of NEI 00-01 Revision 1, instead of NEI 00-01, Revision 2.
3.2.1.1 Attribute Alignment -- Aligns For all of the NEI 00-01, Chapter 3, attributes, the licensee determined that the post-fire safe shutdown analysis aligns directly with the attribute. In these instances, based on the information provided by the licensee in the LAR, as supplemented, and the information provided during the NFPA 805 site audit (that is, the documents reviewed, discussions held with the licensee and the plant tours performed), the NRC staff concludes that the licensee's statements of alignment are acceptable because the analyses are consistent with regulatory guidance for selecting the systems and equipment and their interrelationships necessary to achieve the nuclear safety performance criteria, selection of the cables necessary to achieve the nuclear safety performance criteria, and the identification of the location of nuclear safety equipment and cables.
3.2.1.2 NFPA 805 Nuclear Safety Capability Assessment Methods Conclusion The NRC staff reviewed the documentation provided by the licensee describing the process used to perform the nuclear safety capability assessment required by NFPA 805, Section 2.4.2. The licensee performed this evaluation by comparing the post-fire safe shutdown analysis against the NFPA 805 nuclear safety capability assessment requirements using the NRC-endorsed process in Chapter 3 of NEI 00-01, Revision 2 and documenting the results of the review in LAR Attachment B, Table B-2 in accordance with NEI 04-02, 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 nuclear safety capability assessment with respect to the selection of systems and equipment, selection of cables, and identification of the location of nuclear safety equipment and cables, as required by NFPA 805, Section 2.4.2. The NRC staff concludes that the licensee's method is acceptable because it meets the NRC-er:tdorsed guidance.
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 fire protection programs based on Appendix R to 10 CFR 50 and NUREG-0800, Section 9.5.1, (Reference 78), as well as, in part NEI 00-01, Chapter 3, since NFPA 805 only requires the licensee to maintain the fuel in a safe and stable condition rather than achieve and maintain cold shutdown.
The licensee stated that the NFPA 805 licensing basis is to achieve and maintain Hot Standby, Mode 3, which is the basic safe and stable condition established and maintained for the nuclear safety capability assessment. In SSD RAI 02 (Reference 20), the NRC staff requested additional information to determine if there are any fire areas where it is necessary to perform a transition to cold shutdown. In its response to SSD RAI 02 (Reference 8), the licensee stated that the NSCA currently does not require any fire areas to transition to cold shutdown in order to remain safe and stable. The licensee further stated that in the event the station decides to transition to cold shutdown, the NSCA has conservatively identified the necessary actions to initiate cool down to cold shutdown mode for each fire area. The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated the ability to achieve and maintain basic safe and stable conditions at Hot Standby, Mode 3, and that the ability to proceed to cold shutdown has been maintained.
In SSD RAI 03 (Reference 20), the NRC staff requested additional information regarding instrumentation and controls such as local equipment operation or instrument indications, notpart of the control room evacuation panel (CREP), used to maintain Hot Standby that are not listed as recovery actions in LAR Attachment G. In its response to SSD RAI 03 (Reference 8), the licensee stated that upon evacuation of the main control room, the Primary Control Station (PCS) is shifted from the Main Control Panel to the CREP. The licensee further stated that to transition the PCS to the CREP, actions are required to be performed at the Main Control Panel prior to leaving the Main Control Room, and at the CREP and that in the event that a fire inside the Main Control Room prevents operators from opening all of the quick disconnect switches at the Main Control Panel prior to leaving, the alternate switches in the cable spreading room would need to be opened. The licensee further stated that because the quick disconnects are likely to be operated in the Main Control Room and if not, can be operated on the way to the CREP in the cable spreading room, these actions are not considered recovery actions and that this is the only case where a potential action outside of the PCS is not considered a recovery action. The licensee further stated that the guidance of FAQ 07-0030 was used in the process of establishing recovery actions which states that actions taken to activate, turn on, power up, transfer control or indication, or otherwise enable the PCS are considered taking place at the PCS and are therefore not considered recovery actions. The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated that the only actions outside the Main Control Room not treated as recovery actions are related to transfer to the Primary Control Station at the CREP.
In RAI 13 (Reference 20), the NRC staff requested additional information regarding support systems that are not typically credited (but potentially available) including instrument air, secondary side support, industrial cooling, and other plant systems not associated with a safety function and also to describe how the probabilistic risk assessment (PRA) modeled these systems. In its response to RAI 13 (Reference 8), the licensee stated that Instrument Air was included in the NSCA model and is credited to ensure certain Nuclear Safety Performance Goals are achieved in all areas. The licensee further stated that the FPRA also credits instrument air as a dependency for any components that require instrument air in order to perform a necessary function to support the success criteria of the front line systems modeled in the FPRA and that the system is modeled similar to other systems in the model. The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated that support systems not typically credited have been modeled in the PRA similar to other systems in the model.
In SSD RAI.12 (Reference 20), the NRC staff requested additional information regarding the timing of the modifications installing the new RCP seal materials and insuring the necessary procedures will be in place for seal cooling interruptions until the new seals are fully installed. In response to SSD RAI 12 (Reference 8), the licensee stated that RCP seal materials will be installed and the installation of these seals is expected to extend beyond the implementation of the NFPA 805 Program (180 days following Safety Evaluation receipt). The licensee stated in the LAR that until the new seal materials are installed, procedures for seal cooling interruptions are in place to address the issue as part of the existing Appendix R analysis. In its response to SSD RAI 12 (Reference 8), the licensee further stated that these procedures would continue to be maintained under the approved NFPA 805 program until the planned seal modifications are complete. The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee's procedures to address a loss of RCP seal cooling are in place and will remain in place until the new seals are fully installed.
The licensee's position on safe and stable includes an initial coping/assessment period. The initial coping/assessment period is the period following a receipt of a fire alarm or report of fire to the control room when the operators make initial assessments of the alarm/annunciation/report.
During this period, Operations is determining which fire zone/area is affected based on information coming from the fire detection and control system and local observations in the plant.
Operations is also evaluating the consequences of any other annunciator/alarms that may occur based on potential impacts of fire on cables and/or equipment in that area. The licensee stated that following the initial coping/assessment period at the start of a fire, the operators maintain safe and stable conditions. It is not expected that Operations will perform any safe shutdown operator recovery actions during this period. After the exact fire zone is determined, the results of the initial coping/assessment period will guide operations personnel to obtain the correct procedure to maintain nuclear safety. Once entry into a procedure is determined, then the actions of the procedure are followed.
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 because the licensee conducts an initial coping/assessment period followed by location determination, consequence evaluation, and finally entry i.nto the appropriate procedure to maintain nuclear safety.
3.2.3 Applicability of Feed and Bleed As stated below, 10 CFR 50.48 (c)(2)(iii) limits the use of feed and bleed:
In demonstrating compliance with the performance criteria of Sections 1.5.1 (b) and (c), a high-pressure charging/injection pump coupled with the pressurizer power-operated relief valves (PORVs) as the sole fire-protected safe shutdown path for maintaining reactor coolant inventory, pressure control, and decay heat removal capability (i.e., feed-and-bleed) for PWRs is not permitted.
The NRC staff reviewed LAR Table 5-3, "10 CFR 50.48(c)-Applicability/Compliance Reference,"
and LAR Attachment C, "NEI 04-02 Table B Fire Area Transition," to evaluate whether VCSNS meets the feed and bleed requirements. The licensee stated that feed and bleed is not utilized as the sole fire protected safe shutdown methodology. The NRC staff confirmed this by reviewing the designated safe shutdown path listed in LAR Attachment C for each fire area. The review confirmed that all fire area analyses include the safe shutdown equipment necessary to provide decay heat removal without relying on feed and bleed. In addition the NRC staff confirmed that all fire areas either meet the deterministic requirements of NFPA 805, Section 4.2.3, or the performance-based evaluation performed in accordance with NFPA 805, Section 4.2.4 demonstrates that the integrated assessment of risk, defense-in-depth, and safety margins for the fire area is acceptable. The NRC staff concludes that the licensee meets the requirements of 10 CFR 50.48(c)(2)(iii) because feed and bleed is not utilized as the sole fire-protected safe shutdown methodology.
3.2.4 Assessment of Multiple Spurious Operations NFPA 805 Section 2.4.2.2.1 "Circuits Required in Nuclear Safety Functions" states, in part, that:
Circuits required for the nuclear safety fundions 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 probabilistic safety assessment (PSA) evaluation shall address the risk contribution associated with all potentially risk-significant fire scenariqs. Because the performance-based approach utilized fire risk evaluations in accordance with NFPA 805, Section 4.2.4.2, "Use of Fire Risk Evaluation," adequately identifying and including potential multiple spurious operation (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 (Reference 5), Regulatory Guide 1.205 (Reference 2) and FAQ 07-0038 (Reference 53).
LAR Attachment F states that the licensee conducted an expert panel review composed of a PRA engineer, Operations Engineer, Fire Protection Engineer, Systems Engineer, and an Electrical Engineer. The licensee further stated that the expert panel sources for identifying MSOs included
- 67:..
the safe shutdown analysis, generic lists (provided in NEI 00-01 ), self-assessment results, PRA insights, and operating experience and that the results of the review were integrated into the Fire PRA, the nuclear safety capability assessment (NSCA), and a site-specific calculation that documented the NFPA 805 Multiple Spurious Operations review. The licensee further stated that an evaluation of the MSO's impact on NFPA 805 compliance was performed.
The NRC staff reviewed the licensee's expert panel process for identifying circuits susceptible to multiple spurious operations as described above and concludes that the licensee adopted a systematic and comprehensive process for identifying multiple spurious operations to be analyzed utilizing available industry guidance. The NRC staff also concludes that the process used provides reasonable assurance that the fire risk evaluations appropriately identify and include risk significant multiple spurious operations combinations and that the licensee's approach for assessing the potential for multiple spurious operations combinations is acceptable.
3.2.5 Establishing Recovery Actions NFPA 805, Section 1.6.52, "Recovery Action," defines a recovery action 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.
NFPA 805 Section 4.2.4, "Performance-Based Approach," states, in part, that:
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 (Reference 6) Section 4.2.1.3, "Establishing Recovery Actions," and LAR Attachment G, "Recovery Actions Transition," as supplemented, to evaluate whether the licensee meets the associated requirements for the use of RAs per NFPA 805.
Based on the information provided by the licensee in the LAR, the NRC staff concludes that the licensee's approach for transitioning operator manual actions (OMAs) into the 1 O CFR 50.48(c) risk-informed, performance-based fire protection program as recovery actions is based on NEI 04-02, Revision 2, Section 4.6, "Regulatory Submittal and Transition Documentation," as endorsed with exceptions by RG 1.205, Revision 1 and the guidance of FAQ 07-0030, Revision 5 (Reference 52). The population of OMAs addressed during the NFPA 805 transition process are
- those recovery actions that are required to resolve VFDRs, and are therefore subject to a risk informed evaluation.
As a result of the elimination of the self-induced station blackout (SISBO) compliance strategy, only a limited number of pre-transition OMAs were retained. The remaining recovery and primary control station actions are associated with Control Complex fires, when Control Room evacuation is required. In SSD RAI 03, the NRC staff requested that the licensee describe any instrumentation and controls such as local equipment operation or instrument indications, not part of the control room evacuation panel, used to n;iaintain hot standby that are not listed as recovery actions in LAR Attachment G. In its response to SSD RAI 03 (Reference 8), the licensee stated the following:
"Upon evacuation of the main control room, the Primary Control Station is shifted from the Main Control Panel to the Control Room Evacuation Panel. To transition the PCS to the CREP, actions are required to be performed at the Main Control Panel prior to leaving the_ Main Control Room, and at the CREP.
In the event that a fire inside the Main Control Room prevents operators from
- opening all of the quick disconnect switches at the Main Control Panel prior to leaving, the alternate switches in the cable spreading room would need to be opened. Because the quick disconnects are likely to be operated in the Main Control Room and if not, can be operated on the way to the CREP in the cable spreading room, these actions are not considered Recovery Actions. This is the only case where a potential action outside of the PCS is not GOnsidered a Recovery Action."
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee appropriately identified the credited nuclear safety capability assessment primary control station and recovery actions required for a control room evacuation.
LAR Section 4.2.1.3 and LAR Attachment G discuss the methodology and results associated with*
treatment of OMAs. VCSNS has no recovery actions other than those associated with the Control Complex fires, when Control Room evacuation is required. For fires outside the control complex, it was determined that no recovery actions were required in order to meet the criteria of NFPA 805 Section 4.2.4, either for achieving risk goals or for meeting defense-in-depth.
OMAs meeting the definition of a recovery action are required to comply with the. NFPA 805 requirements outlined above. Some of these 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 the risk-informed, performance-based fire protection program because of the defense-in-depth considerations described in NFPA 805, Section 1.2. Based on a review of the LAR, the NRC staff concludes that there are no defense-in-depth recovery actions required for NFPA 805 transition.
The licensee stated that all recovery actions listed in LAR Attachment G were subjected to a feasibility review and that recovery actions were evaluated against the feasibility criteria provided in the NEI 04-02, FAQ 07-00_30, and RG 1.205. The basis and the results of the assessment are included in a site-specific calculation and LAR Attachment G describes the process for evaluating recovery action feasibility and lists the individual feasibility attributes used in the assessment process. LAR Attachment G, Table G-1, "Recovery Actions and Activities Occurring at the Primary Control Stations," describes each recovery action associated with disposition of a VFDR from the fire area assessments as documented in LAR Attachment C, "Fire Area Transition." As stated in the LAR, the licensee assessed the recovery actions listed in Table G-1 to the feasibility criteria in FAQ 07-0030.
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 recovery actions in accordance with NFPA 805, thereby providing reasonable assurance that the regulatory requirements of 10 CFR 50.48(c) will be met. The NRC staff concludes that the feasibility criteria applied to recovery actions are acceptable based on conformance with the endorsed guidance contained in NEI 04-02.
3.2.6 Plant-Specific Treatments or Technologies (Incipient Fire Detection Systems)
The licensee proposed the installation of an incipient fire detection system to improve early indications of fire precursors in certain risk significant areas of the plant, specifically inside key electrical panels.
As described in LAR Attachment C, a modification to install an incipient detection system is credited with reducing risk in fire areas CB06 "CB Relay Room" and CB15 "CB Upper Cable Spreading Room". LAR Attachment S, Table S-1, Item ECR50811, is to install incipient detection at the top of selected electrical panels in the Relay and Upper Cable spreading rooms. In a letter dated December 11, 2014 (Reference 19), the licensee revised LAR Attachment S, included a Table S-3 for completed modifications and included Item ECR50811 in Table S-3,. indicating that the modification to install incipient detection has been completed in 2013.
In FPE RAI 01 (Reference 20), the NRC staff requested that the licensee provide more details about the incipient detection system. Specifically, the NRC staff requested the licensee provide more details regarding system design features, NFPA code(s) of record, installation, acceptance testing, setpoint control, alarm response procedures and training, and routine inspection, testing, and maintenance. In its response to FPE RAI 01 (Reference 8), the licensee stated that FAQ 08-0046 (Reference 57), EPRI 1016735 (Reference 79), and NUREG/CR-6850, Supplement 1 (Reference 43) will be used as design inputs in system selection, design, and installation for protection of Relay Room and Upper Cable Spreading Room Panels as described in the Fire PRA and that this includes design features, acceptance testing, and routine inspection testing and maintenance. The licensee further stated that in accordance with FAQ 08-0046, NFPA 76-2009 (Reference 66), "Standard for the Fire Protection of Telecommunications Facilities" will be used as the code of record for the design of the system. The licensee further stated that deviations from these documents for the system installation are not expected and that a code compliance document will be developed and maintained. The licensee further stated that sensitivity settings will meet those defined in the November 23, 2009, FAQ 08-0046 Closure Memorandum and that regular functional testing and maintenance will be performed in accordance with NFPA codes and
- applicable vendor recommendations and that the systems will be included within the overall NFPA 805 Monitoring Program.
The NRC staff concludes that the licensee's response to the RAI and the description of the installation and maintenance aspects related to the proposed installation of incipient detection are acceptable beca.use:
The installation of the system will be performed in accordance with the appropriate NFPA codes, the equipment manufacturers' requirements, and the guidance in FAQ 08-0046, "Incipient Fire Detection Systems";
The configuration of the system will be such that the sensitivity settings will meet the settings defined in FAQ 08-0046; The equipment will be maintained and functionally tested in accordance with the vendor recommendations aswell as the guidance contained in FAQ 08-0046; The licensee identified a modification in LAR Attachment S, Table S-1, to install the system, which would be required by the proposed license condition.
As stated previously, in a letter dated December 11, 2014 (Reference 19), the licensee revised LAR Attachment S, included a Table S-3 for completed modifications and included Item ECR50811 in Table S-3, indicating that the modification to install incipient detection has been completed in 2013.
3.2.7 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 nuclear safety capability assessment. The NRC staff concludes that the fuel safe and stable condition, proposed by the licensee, is acceptable because the licensee's process provides reasonable assurance that it will appropriately identify and locate the systems, equipment, and cables required to achieve and maintain the fuel in a safe and stable condition, as well as to meet the NFPA 805 nuclear safety performance criteria.
- The NRC staff confirmed, through review of the documentation provided in the LAR, that feed and bleed was not the sole fire-protected safe shutdown path for maintaining reactor coolant inventory, pressure control, and decay heat removal capability, in accordance with 10 CFR 50.48(c)(2)(iii).
The NRC staff reviewed the licensee's process to identify and analyze multiple spurious operations. Based on the information provided in the LAR, as supplemented, the NRC staff concludes that the process used to identify and analyze multiple spurious operations is comprehensive and thorough. The NRC staff noted that through the use of an expert panel process in accordance with RG 1.205, NEI 04-02 and FAQ 07-0038, potential multiple spurious operation combinations were identified and included as necessary into the nuclear safety capability assessment as well as the applicable fire risk evaluations. The NRC staff considers the licensee's approach for assessing the potential for multiple spurious operation combinations to be acceptable because it was performed in accordance with NRG-endorsed guidance.
The NRC staff concludes that the process used by the licensee to review, categorize and address recovery actions during the transition is consistent with the NRG-endorsed guidance contained in NEI 04-02 and RG 1.205, and therefore, the regulatory requirements of 10 CFR 50.48(c) and NFPA 805 for nuclear safety capability assessment methods are met.
The NRC staff reviewed -the proposed installation of an incipient detection system to monitor conditions in certain electrical cabinets. Based on the information provided in the LAR, as supplemented, the NRC staff concludes that the fire protection aspects regarding installation, testing, and operation of the proposed incipient detection system are acceptable because the installation, testing, and operation of the system will be done in accordance with appropriate NFPA codes, original equipment manufacturer requirements, NUREG/CR-6850, Supplement 1, NRC FAQ 08-0046, and the licensees preventative maintenance program requirements and because the licensee included the action to install the system in LAR Attachment S, Table S-2, which would be required by the proposed license condition.
3.3 Fire Modeling Performance-Based Approach NFPA 805 (Reference 1) allows both fire modeling and fire risk evaluations as performance-based alternatives to the deterministic approach outlined in the standard. These two performance-based approaches are described in NFPA 805, Sections 4.2.4.1 and 4.2.4.2, respectively. Although fire modeling and fire risk evaluations are presented as two different approaches for performance-based compliance, the fire risk evaluation approach generally involves some degree of fire modeling to support engineering analyses and 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 6) Section 4.5.2, "Performance-Based Approaches,
which describes how the licensee used fire modeling as part of the transition to NFPA 805 at VCSNS, and LAR Section 4.7.3, "NFPA 805 Quality Requirements (NFPA 805, Section 2.7.3),
which describes how the licensee performed fire modeling calculations in compliance with the NFPA 805 performance-based evaluation quality requirements for fire protection systems and features at VCSNS, to determine whether the fire modeling used to support transition to NFPA 805 is acceptable.
In LAR Section 4.5.2, the licensee indicated that the fire modeling performance-based approach was used to support compliance with NFPA 805, Section 4.2.4.1. The results of the NRC staff review of the technical adequacy of fire modeling performed at VCSNS in support of compliance with NFPA 805, Section 4.2.4.1, are discussed below. The licensee also used the fire risk evaluation performance-based method per NFPA 805, Section 4.2.4.2 (i.e., fire PRA), with input from fire modeling analyses. Therefore, the NRC staff reviewed the technical adequacy of the VCSNS fire risk evaluations, including the supporting fire modeling analyses, as documented in Section 3.4.2 of this safety evaluation, to evaluate compliance with the nuclear safety performance criteria.
NFPA 805, Section 2.4.1, "Fire Modeling Calculations, specifically addresses the application requirements for using PB fire models as follows:
NFPA 805, Section 2.4.1.2.1, "Acceptable Models," states that Only fire models that are acceptable to the authority having jurisdiction shall be used in fire modeling calculations.
NFPA 805, Section 2.4.1.2.2, "Limitations of Use," states that:
Fire models shall only be applied within the limitations of that fire model.
NFPA 805, Section 2.4.1.2.3, "Validation of Models," states that:
The fire models shall be verified and validated.
NFPA 805, Section 4.2.4.1, "Use of Fire Modeling," identifies the approach for use of fire modeling as a performance-based method, including: identify targets, establish damage thresholds, determine limiting condition(s), and establish fire scenarios, protection of required nuclear safety success path(s), and operations guidance.
In addition, RG 1.205, Revision 1 (Reference 2), Regulatory Position C.4.2, and NEI 04-02, Revision 2 (Reference 5), Section 5.1.2, "Fire Modeling Considerations," provide guidance by identifying fire models that the NRC staff considers acceptable for use by plants transitioning to a risk-informed, performance-based fire protection program in accordance with NFPA 805 and 10 CFR 50.48(c).
3.3.1 Overview of the Fire Modeling Performance-Based Approach The fire modeling performance-based approach was used to address variances from deterministic requirements (VFDRs) in the following fire areas:
Fire Area CB10: East intermediate cable chase Fire Area CB12: Northeast intermediate cable chase Fire Area CB18: East main floor cable chase Fire Area 1811: Service water booster pump cooling unit room The performance-based approach applied in these fire areas involved the following steps:
- 1.
Identify Targets: Three VFDRs were identified in each of fire areas CB10, CB12, and 1811. Four VFDRs were identified in fire area CB18. Each VFDR is associated with a specific set of cables, which were considered targets in the fire modeling analysis.
- 2.
Establish Damage Thresholds: The cable targets were assumed to fail when the applicable critical temperature or heat flux specified in NUREG/CR-6850 is.
reached.
- 3.
Determine Limiting Conditions: This involved identifying the combination of targets with the highest susceptibility to damage in any fire environment, i.e.,
immersion in a flame or plume, radiant exposure, or hot gas layer (HGL) immersion.
- 4.
Establish Fire Scenarios: Based on the types of combustibles that can be present, transient fires were considered in the four fire areas. In addition, an electric motor fire scenario was postulated in fire area IB11. The maximum expected fire scenario (MEFS) was defined for each fire area and ignition source as the most challenging fire that could reasonably be anticipated in the area. The limiting fire scenario (LFS) was then determined by varying (increasing) the heat release rate (HRR) of the fire to determine the threshold at which a target would exceed the critical temperature or radiant heat flux.
- 5.
Fire Protection of Required Nuclear Safety Success Paths: Protection of the required nuclear safety success paths is achieved by fire prevention (e.g.,
transient combustible controls) and mitigation (e.g., automatic detection and suppression).
- 6.
Operations Guidance: Based on the nature and low frequency of operator
- activities in fire areas CB10, CB12, and CB18, no specific operational guidance pertaining to these areas is provided to plant personnel. In fire area IB11, plant personnel must follow existing administrative procedures to ensure that the fire modeling assumptions are not violated.
3.3.2 Defense-in-Depth In addition to the fire modeling evaluations, the following elements contribute to achieving defense-in-depth in the fire areas where the. fire modeling performance-based approach was performed:
Fire Prevention: The licensee stated that no transient combustibles will be allowed in areas where the performance-based fire modeling approach was used without further evaluation and approval to ensure that fire damage is prevented as documented in the fire model.
Fire Protection: The licensee stated that fire zones CB10, CB12, and CB18 have automatic detection and suppression and fire zone IB11 has automatic detection which is not taken into consideration in the performance-based fire modeling calculations.
3.3.3 Safety Margins Safety margins are quantified based on the MEFS and LFS and defined in terms of: (1) The increase of the HRR above that in the MEFS, required to create a HGL temperature that will result in target damage; and, (2) The distances from the postulated ignition source that need to be maintained to prevent fire propagation to cable tray targets. In order to account for uncertainties and unknowns in the analytical process and to ensure adequate defense-in-depth, the LFS must be sufficiently greater than the MEFS.
The licensee stated that the margin between the peak heat release rate for the MEFS and LFS varied from 80% to 500% and that the margin between the mass of the combustible material for,
the MEFS and LFS varied from 78% to 400%. The licensee's analysis assumed a fuel composition of polystyrene and consid~red the impact of spatial factors, such as wall and corner effects, on the heat release rate, which represents a conservative and bounding step in the analysis. The presence of smoke detectors in the subject areas also provides a margin of defense-in-depth in the licensee's analysis. The NRC staff concludes that the safety margin provided in the licensee's analysis is acceptable because the licensee's assumptions in the analysis are conservative.
3.3.4 Fire Models Used in the Analysis The following algebraic fire models and correlations were used in determining the staging area for transient fire sources and the heat release rate of the LFS:
Heskestad's Plume Temperature Correlation Point Source Radiation Model These algebraic models are described in NUREG-1805, "Fire Dynamics Tools (FDTs):
Quantitative Fire Hazard Analysis Methods for the US Nuclear Regulatory Commission Fire Protection Inspection Program" (Reference 47). 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 48).
The Consolidated Model of Fire and Smoke Transport (CFAST) computational fire model, Version 6 was used to determine the potential for the generation of an HGL. The V&V of CFAST is documented in NUREG-1824, Volume 5 (Reference 48).
The V&V of the fire models that were used in the transition to NFPA 805, including fire models used to support compliance with NFPA 805 Section 4.2.4.1, are discussed in more detail in SE Section 3.8.3.2.
For each fire area and corresponding MEFS, the licensee either demonstrated that the model input parameters were within the validated range or performed a sensitivity analysis to justify application of the model outside the limitations of its use which is consistent with the approach described in NUREG-1934, "Nuclear Power Plant Fire Modeling Analysis Guidelines,"
(Reference 81 ).
3.3.5 RAls Pertaining to the Performance-Based Fire Modeling Approach In letters dated July 26, 2012 (Reference 20), and August 13, 2013 (Reference 21 ), the NRC staff requested additional information from the licensee. In letters dated October 10, 2012 (Reference 8), February 1, 2013 (Reference 9), April 1, 2013(Reference10), and October 14, 2013 (Reference 11 ), the licensee provided its responses to the requests.
During the onsite audit, the NRC staff noted that the licensee's performance-based fire modeling of transient fire scenarios assumes that no combustibles be placed within one foot of the adjacent walls, so that wall effects of the fire source do not have to be accounted for. In fire modeling (FM) RAI 01 (s) (Reference 20), the NRC staff requested that the licensee provide technical justification for using the separation distance of one foot.
In its response to FM RAI 01 (s) (Reference 8), the licensee stated that a sensitivity case was investigated for each of the performance-based fire modeling reports in which the edge of the transient staging area was located at one foot from a wall or corner. The licensee further stated a fire was simulated by increasing the following inputs by a factor of two for a wall effect or a factor of four for the corner effect: the fire size, the fire enclosure volume, the fire enclosure wall and ceiling areas, and the natural ventilation area. The licensee further stated the results show that by using th~ image method to determine the impacts of a wall or corner fire, the peak temperature in the room is still below the damage temperature of 205 °C.
In FM RAI 01.03 (Reference 21 ), the NRC staff requested that the licensee address the wall or corner effect when a fire is within 2 feet of a wall. In its response to FM RAI 01.03 (Reference 11 ), the licensee stated that the performance based fire modeling reports for several fire areas were revised to include the wall effects as the base case for fire modeling since the transient staging areas are located one foot from the wall.
The NRC staff concludes that the licensee's responses to the RAls are acceptable because the licensee demonstrated that the peak temperature in the room is below the damage threshold and because the licensee revised their performance-based fire modeling reports to account for wall effects of the fire source.
During the onsite audit, the NRC staff and the licensee viewed fire area CB 10 and the NRC staff observed two insulated copper pipes in the corner of the area closest to the transient combustible staging area. In FM RAI 01.u (Reference 20), the NRC staff requested that the licensee explain why the observed pipe insulation was not considered in the performance-based fire modeling analysis for fire area CB10.
In its response to FM RAI 01.u (Reference 8), the licensee explained that the pipes are insulated with fiberglass, jacketed with a vapor barrier laminate of aluminum foil and glass cloth with lap adhesive, and.that the insulation meets the specifications in American Society of Testing Materials ASTM C547.
In FM RAI 01.04 (Reference 21) the NRC staff requested that the licensee provide the flame spread index and the smoke developed index for the pipe insulation. In its response to FM RAI 01.04 (Refer.ence 11 ), the licensee stated that it found that the pipe insulation does not exceed 25 flame spread index, 50 smoke developed index when tested in accordance with ASTM E84.
In its response to subsequent FM RAI 09.02 (Reference 11 ), the licensee stated that the administrative controls (procedure) will identify the fire area as a restrictive combustible zone and that no transient combustibles will be allowed in the area without further evaluation and approval to ensure that fire damage is prevented as documented in the fire model. The licensee included an action to revise its fire
I protection program administrative procedures (e.g. FP Program Plan, Transient Material Control, Compensatory Measures) as needed for implementation of the NFPA 805 Program in LAR Attachment S, Table S-2, Implementation Item 1 and the NRC staff concludes that this action is acceptable because it will result in compliance with NFPA 805 and would be required by the proposed license condition.
The NRC staff concludes that the licensee's basis for not considering the observed pipe insulation as an intervening combustible in the analysis performed for CB 10 is acceptable because the licensee justified the assumed non-combustible nature of the insulation material and will identify the area as a restrictive combustible zone in its revision to its administrative controls identified in LAR Attachment S, Table S-2, Implementation Item 1 to not allow transient combustibles in the area without appropriate evaluation of the performance-based fire model.
During the onsite audit, the NRC staff and the licensee viewed fire area 1811 and the NRC staff observed several exposed cables extending from the tray that contains the VFDR cables toward the proposed transient staging area on the floor.
The NRC staff noted that the licensee's analysis used the distance to the staging area from the edge of the VFDR cable tray, rather than the exposed cable. In FM RAI 01.v (Reference 20), the NRC staff requested that the licensee quantify the effect of the exposed cable on the conclusions of the analysis.
In its response to FM RAI 01.v (Reference 8), the licensee stated that it moved the transient staging area 2 feet further away from the VFDR tray, thus increasing the distance between the edge of the transient staging area and the exposed cable targets. The licensee further stated that the revised configuration has no impact on the results and conclusions of the analysis in that the revised analysis did not result in cable damage for a fire originating in the transient staging area.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee revised its analysis to accurately reflect the location of the exposed cable and demonstrated that doing so has no impact on the results of the analysis.
In FM RAI 08 (Reference 20), the NRC staff requested that the licensee discuss the uncertainties and safety margin that exist for fire areas utilizing the fire modeling performance-based approach, and to provide a technical justification for how the analysis meets the regulatory requirements.
In its response to FM RAI 08 (Reference 8), the licensee described how the uncertainties in the fire properties (i.e., the location of the fire, quantity of combustible material, and HRR of the transient material stored) were addressed and also discussed the safety margins associated with the parameters used in the performance-based fire modeling approach, such as the fire properties, model bias of CFAST, the presence (or not) of a fixed ignition source, smoke detection in the compartment, and the time allowed for the fire brigade to arrive.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee provided technical justification for the uncertainties and safety margin that exist and demonstrated compliance with NFPA 805 Section 4.2.4.1.5 for the fire areas that utilize the performance-based approach.
During the NFPA 805 site audit, the NRC staff found that the licensee's performance-based fire modeling analyses depended on limiting combustible material size, type, and location. The licensee explained that it planned to mark a transient staging area on the floor in the rooms where combustible material would be allowed to be stored.
In FM RAI 09 (Reference 20), the NRC staff requested that the licensee discuss implementation of post-transition procedures that will maintain a program that is consistent with the fire modeling calculations, and identify features or circumstances that would ensure safety in the event that administrative controls are not followed in the fire areas where the performance-based fire modeling approach was performed.
In its response to FM RAI 09 (Reference 8), the licensee stated that the process of developing applicable procedures will be ba&ed in part on the results and insights
- generated by the fire modeling studies, which were discussed in detail with the NRC staff during the audit. The licensee further stated that the defense-in-depth in the fire areas where the performance-based fire modeling approach was used was addressed by including sensitivity calculations for various parameters in the respective fire modeling reports. The licensee included an action to complete the development and issuance of the Fire Safety Analysis to summarize area results and insights from the NFPA 805 analysis in LAR Attachment S, Table S-2, Implementation Item 18, and the NRC staff considers this action acceptable because it will result in compliance with NFPA 805 and would be required by the proposed license condition.
In FM RAI 09.01 (Reference 21 ), the NRC staff requested that the licensee provide a description of how the technical results of the fire modeling studies will be translated into administrative controls that will be used in the plant.
In its response to FM ~I 09.01 (Reference 11 ), the licensee described how the results of the performance-based fire modeling studies will be translated into administrative controls, the training that will be provided to operations and maintenance personnel regarding these administrative controls, and the process being proposed to verify compliance with the administrative controls. The licensee stated that for the fire zones where the fire modeling performance-based approach was used as the method of compliance to NFPA 805, under normal operations, combustible storage will not be allowed in the fire zone and that if under special circumstances, temporary combustible storage would be required, the details in the performance-based fire modeling reports provide fire modeling insights to allow contingency planning for exceptions to the ban on combustible storage. The licensee further explained how the proposed approach avoids over-reliance on performing programmatic activities based on the guidance in NEI 04-02 (Reference 5) and that training on transient combustible controls is addressed by Implementation Items 1 and 15 in LAR Attachment S, Table S-2.
In FM RAI 09.02 (Reference 21 ), the NRC staff requested that the licensee provide justification for implementing an administrative controls program that deviates from the standard of practice used throughout the nuclear industry. In its response to FM RAI 09.02 (Reference 11 ), the licensee stated that the administrative controls (procedure) will identify the entire fire area(s) as a restrictive combustible zone(s) and that no transient combustibles will be allowed in the area without further evaluation and approval to ensure that fire damage is prevented as documented in the fire model. The licensee included an action to revise its fire protection program administrative procedures (e.g. FP Program Plan, Transient Material Control, Compensatory Measures) as needed for implementation of NFPA 805 Program in LAR Attachment S, Table S-2, Implementation Item 1 and the NRC staff concludes that this action is acceptable because it will result in compliance with NFPA 805 and would be required by the proposed license condition.
The NRC staff concludes that the licensee's response to the RAls is acceptable because the licensee demonstrated that the nuclear safety and radioactive release performance criteria, goals, and objectives will continue to be met in case of violations of the administrative controls assumed in the performance-based fire modeling calculations and because the licensee identified required actions that will incorporate the provisions of NFPA805 in the licensee's fire protection program and included those actions as implementation items in LAR Attachment S, which would be required by the proposed license condition.
3.3.6 Conclusion for Section 3.3 The NRC staff reviewed the licensee's performance-based fire modeling analyses and the RAI responses. The NRC staff concludes that, subject to completion of the implementation items as stated in the proposed license condition, the fire models used to support compliance with NFPA 805 Section 4.2.4.1 are acceptable, meet the V&V requirements, and are applied within their limitations of use. The NRC staff also concludes that the licensee's analyses provides adequate justification that the cables identified as VFDR targets in fire areas CB 10, CB 12, CB 18 and IB11 will not be damaged by the MEFSs and that adequate safety margin exists between the MEFSs and LFSs.
3.4 Fire Risk Evaluations 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 both the fire modeling and fire risk evaluation performance-based methods in accordance with NFPA 805 Sections 4.2.4.1 and 4.2.4.2, respectively. Implementation of the fire risk evaluation performance-based method is discussed in this section. Implementation of the fire modeling performance-based method is discussed in Section 3.3.
NFPA 805, Section 4.2.4.2, "Use of Fire Risk Evaluations," states that:
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.
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 risk shall be calculated using the approach described in [NFPA 805, Section 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 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:
Preventing fires from starting; Detecting fires quickly and extinguishing those that do occur, thereby limiting fire damage; and 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 4.5.2.2, "Fire Risk Approach," and LAR Attachment C Table B-3, "C. NEI 04-02 Table B-3 Fire Area Transition," as supplemented by the licensee's letter dated February 25, 2014 (Reference 14).
When implementing the performance-based approach, the licensee followed the guidance contained in NEI 04-02, Section 5.3, "Plant Change Process," which includes a detailed consideration of defense-in-depth and safety margins as part of the change process. The license documented the method used to meet the DID requirements of NFPA 805 in LAR Attachment C Table B-3, as supplemented by the license's letter dated February 25, 2014 (Reference 14). LAR Attachment C, Table B-3 and LAR Tables 4-9 and 4-10 document the results of.the licensee's review of fire suppression and fire detection systems.
The licensee's methodology for evaluating DID refers to each of the three DID elements identified in NFPA 805, Section 1.2 as Echelons 1, 2, and 3, respectively. As described in the response to PRA RAI 05 (Reference 8), this method was implemented in the fire risk evaluations (FREs) performed on applicable performance-based fire areas~ The licensee identified the types of fire protection systems and features that addressed the echelons.
In the DID evaluation, the licensee identified the fire protection systems and features credited for the echelons and qualitatively assessed if the credited systems and features provided an adequate balance between the echelons. Additional fire protection systems and features were credited if necessary to improve the balance between the echelons.
Specifically with regard to Echelons 1 and 2, the licensee explained that the evaluation of these echelons considered high frequency fire scenarios, defined as scenarios that have a frequency higher that 0.1 per year, in determining if additional fire protection systems and features were needed to provide an adequate balance. With regard to Echelon 3, the licensee explained that the evaluation of this echelon considered high-consequence fire scenarios, defined as scenarios having a core damage frequency (CDF) of 1 E-06 per reactor-year or greater and a conditional core damage probability (CCDP) of 0.1 or greater, in determining if additional fire protection systems and features were needed to provide an adequate balance. The licensee's conclusion from the DID assessments was that no actions or changes to the plant or procedures was needed to maintain the philosophy of DID. The results of the licensee's DID assessment for the echelons by fire area is provided in LAR Attachment C, Table B-3, as supplemented by the licensee's letter dated February 25, 2014 (Reference 14).
Based on its review of the LAR and the response to PRA RAI 05 (Reference 8), and the FREs during its audit of the VCSNS 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. The NRC staff also 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," discusses how safety margins are addressed as part of the FRE process and that this process is based on the requirements of NFPA 805, industry guidance in NEI 04-02, and RG 1.205. An FRE was performed for each fire area containing VFDRs for which the fire modeling performance-based method, NFPA 805 Section 4.2.4.1, was not used. The FREs contain the details of the licensee's review of safety margins for each applicable performance-based fire area.
LAR Section 4.5.1.2 states that the Fire PRA applies methodologies consistent with the guidance in NUREG/CR-6850 and fire PRA-related frequently asked questions (FAQs) and that fire models acceptable to the NRC were used in the development of the Fire PRA. In its response to PRA RAI 05 (Reference 8), the licensee further described the methodology used to evaluate safety margins in the FREs to include the following elements:
Fire Modeling: Fire modeling performed to support the FREs utilized codes and standards developed by NRC and industry staff (e.g., heat release rates used in the Fire PRA were based on NUREG/CR-6850).
Plant System Performance: Plant system performance parameters were not modified as a result of the FREs.
PRA Logic Model: The Fire PRA model was developed in accordance with industry codes and standards. Specifically, according to LAR Section 4.5.1.2, the Fire PRA was developed in accordance with the ASME/ANS RA-Sa-2009 PRA standard (Reference 36) and, according to the response to PRA RAI 11 (Reference 8), in accordance with RG 1.200, Revision 2 (Reference 35).
The licensee concluded from its safety margin evaluation that no actions or changes to the plant or procedures were needed to maintain sufficient safety margins. The res.ults of the licensee's safety margin assessment by fire area are provided in LAR Attachment C, Table B-3, as supplemented by the licensee's letter dated February 25, 2014 (Reference 14).
The safety margin cr~eria described in NEI 04-02, Section 5.3.5.3 and the LAR, as supplemented, are consistent with the criteria as described in RG 1.17 4 and therefore acceptable. The licensee used appropriate codes and standards (or NRC guidance), and met the safety analyses acceptance criteria in the licensing basis. Based on its review of the LAR and the FREs during its-audit of the licensee's NFPA 805 transition RI/PB FPP, the NRC staff concludes that the licensee's approach adequately addressed the issue of safety margins in the implementation of the fire risk evaluation process.
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 LAR Section 4.5.1, "Fire PRA Development and Assessment," LAR Section 4.7, "Program Documentation, Configuration Control, and Quality Assurance," LAR Attachment C, "C. NEI 04-02 Table B-3 Fire Area Transition," as supplemented by the licensee's letter dated February 25, 2014 (Reference 14),
LAR Attachment U, "Internal Events PRA Quality," LAR Attachment V, "Fire PRA Quality, and LAR Attachment W, "Fire PRA Insights," as supplemented by the licensee's letter dated May 2, 2014Property "Letter" (as page type) with input value "RC-14-0067, License Amendment Request - LAR-06-00055, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. (Reference 15), as well as associated supplemental information.
The licensee initially developed its internal events PRA prior to 2002 when the consensus American Society of Mechanical Engineers (ASME) PRA standard (Reference 36) was first issued 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" (Reference 35) 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 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.
r 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 SE Section 3.8.3, the licensee has a program for ensuring that developers and users of these models are appropriately trained and qualified. Therefore the NRC staff concludes that the PRA should be capable of supporting post-transition Plant Change Evaluations 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 consisted of a full scope-peer review by the Westinghouse Owner's Group 0fVOG) performed in August 2002 using the NEI 00-02 (Reference 82) industry PRA peer review process. Subsequent gap assessments were performed in April 2005 to determine the scope of work required to ensure the PRA meets RG 1.200, Revision 1 (Reference 83) and in November 2007 against RG 1.200, Revision 1. In response to PRA RAI 11 (Reference 8) the licensee stated that an additional gap assessment was performed against RG 1.200, Revision 2 (Reference 35) and that the results of this assessment did not significantly affect the fire PRA results. In response to PRA RAI 80 (Reference 8) the licensee confirmed that the internal events PRA model that incorporated changes made to disposition the facts and observations (F&Os) from these reviews is the model that was used in the development of the FPRA.
For many Supporting Requirements (SRs), there are three degrees of "satisfaction" referred to as Capability Categories (i.e., I, II, and Ill)'. with I being the minimum, II considered widely acceptable, and Ill indicating the state of the art. For other SRs the Capability Categories may be combined (e.g., the requirement for meeting Capability Category I may be combined with II) or the requirement may be the same across all Capability Categories so that the requirement is simply met or not met. For each SR, the PRA reviewer from the peer review team designates one of the Capability Categories or indicates that the SR is met or not met.
LAR Attachment U, Table U-1 of the LAR provides the licensee's resolution of all 23 Level A (important and necessary to address before the next regular PRA update) and Level 8 (important and necessary to address, but resolution may be deferred until the next PRA update) F&Os from the 2002 WOG peer review and the 20 F&Os from the 2005 and 2007 gap assessments. The licensee did not identify any additional F&Os in response to the gap assessment discussed in the response to PRA RAI 11 (Reference 8). In general, an F&O is written for any SR that is judged not to be met or does not fully satisfy Capability. Category 11 of the ASME standard, consistent with RG 1.200.
(
As described in LAR Attachment U, the licensee resolved 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 resolutions. The NRC staff evaluated each F&O and the licensee's resolution in LAR Attachment U to determine whether the F&O had any significant impact for the application. The NRC staff's review and conclusions for resolution of each F&O is summarized in the NRC's Record of Review dated November 7, 2014 (Reference 84).
Additional information associated with F&O L2-02 with respect to exclusion of early containment overpressure failures from the LERF model was requested in PRA RAI 54 (Reference 20 and 22).
In response to PRA RAI 54 (Reference 8 and 12) the licensee stated that the internal events PRA Level 2 model is currently being upgraded to, among other things, improve the LERF model to explicitly treat contributors to potential early containment overpressure failures such as hydrogen burn and direct containment heating and that, while the LERF model update has been completed, it has not been incorporated into the FPRA model. However, the licensee stated that the new model will not result in any new risk insights or significant changes in the risk results because LERF is dominated by containment bypass (which is treated essentially the same in the FPRA model) and steam generator tube rupture (SGTR) sequences (which do not contribute to the FPRA risk results). The NRC staff finds this acceptable because the LERF model update will have no effect on the fire PRA.
In PRA RAls 64 (Reference 20) and 64.01 (Reference 26), regarding the resolution to F&O IE-01-Ga, the NRC staff requested additional information on how the PRA was revised to address the identified weaknesses in the interfacing systems loss of coolant accident (ISLOCA) model and on an assessment of the relevance of more recent data on MOV rupture failure rate to the internal events PRA results. In response to these RAls (References 8 and 17) the licensee explained that the ISLOCA model has been updated since the gap assessment to explicitly model ISLOCA failure modes within the fault tree. The licensee further explained that these failure modes were represented by basic events (component failure probabilities) that are assigned error factors, that basic events are correlated, and that the updated model explicitly models recovery for ISLOCA flow paths in the ISLOCA event tree. Using the updated model, the licensee reported an estimate of the mean CDF (4.15E-06 per year) and LERF (1.09E-07 per year) for internal events and, concluded that the RG 1.17 4 risk guidelines continue to be met. Based on the results of the licensee's uncertainty analysis, which utilized an updated ISLOCA model that explicitly accounts for ISLOCA component failure probabilities that are correlated, the NRC staff concludes that the licensee's conclusion is acceptable and that the RG 1.17 4 risk guidelines are met and that this issue is resolved for this application.
As a result of the review of the LAR and responses to PRA RAls, the NRC staff concludes that the internal events PRA is 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 or has no significant impact on the Fire PRA. Accordingly, the NRC staff concludes that the licensee demonstrated that the internal events PRA meets the guidance in RG 1.200, Revision 2, that it is reviewed against the applicable supporting requirements 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 Fire PRA model by conducting a full-scope peer review of the Fire PRA model using the NEI 07-12 process (Reference 85), and the*
combined PRA standard, ASME/ANS-RA-Sa-2009 (Reference 36), as clarified by RG 1.200, Revision 2 (Reference 35). The full scope peer review of the Fire PRA was performed in August 2010. In addition, a follow-on peer review was performed in February 2011 to address SRs not reviewed during the original peer review due to the associated technical elements not being sufficiently completed for review. These peer reviews serve as the basis for the quantitative risk evaluations for the LAR.
LAR Attachment V, Table V-18 provides the licensee's resolution of all 42 Finding-level F&Os, per peer review guidelines, written against the SRs of the combined PRA standard, ASME/ANS RA-Sa-2009 fire PRA standard (Reference 36) as clarified by RG 1.200, Revision 2 (Reference 35). In response to PRA RAI 77 (Reference 8), the licensee clarified that all F&Os from the follow-on peer review described in LAR Attachment V, Table V-18 as "DRAFT" were confirmed to be the final F&Os from the peer review. LAR Attachment V, Table V-16 identifies all SRs that were determined by the peer review to be met only at CC-I. An F&O was written against each SR determined to only meet CC-I, and the F&O and its resolution either provided in the LAR or in response to NRC staff RAls.
In PRA RAI 14 (Reference 20), the NRC staff requested that the licensee identify any changes made to the Fire PRA that are consistent with the definition of a "PRA upgrade" since the last full-scope peer review of PRA models as defined by the ASME/ANS PRA Standard (Reference 36). In its response to PRA RAI 14 (Reference 8), the licensee stated that no changes meeting the definition of a PRA upgrade were made to the Fire PRA since the peer review.
The NRC staff reviewed the licensee's resolution of all of the F&Os to determine the technical adequacy of the Fire PRA for the NFPA 805 application. The NRC staff's review and conclusion for the licensee's resolution of each of the F&Os is summarized in in the NRC's Record of Review dated November 7, 2014 (Reference 84). The NRC staff requested additional information in support of its review of the F&Os, as discussed below.
Regarding the lack of plant procedures to address seismic-induced fires noted in F&O SF~A4-01, the licensee explained in response to PRA RAI 35 (Reference 8) that FPRA documentation has been revised to provide recommendations to address seismic issues in plant procedures and training. Furthermore, the licensee described an action in Implementation Item 21 of LAR Table S-2, as supplemented by the licensee's letter dated November 26, 2013Property "Letter" (as page type) with input value "RC-13-0166, License Amendment Request - LAR-06-00055, License Amendment Request to Adopt NFPA 805, Response to Request for Additional Information" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. (Reference 12), to incorporate the seismic-fire interaction recommendations into plant procedures and training. The NRC staff finds that seismic-induced fire evaluation is acceptable because the procedures have been evaluated to support transition and the results of this evaluation will be incorporated into the appropriate procedures after transition through an implementation item which would be required by the proposed license condition.
In PRA RAI 68 (Reference 20), the NRC staff requested that the licensee provide justification that the transient zone boundary appropriately accounts for fire propagation to adjacent transient zones due to flame spread along cables or due to increased heat release rate (HRR) from secondary combustibles. In response to the RAI (Reference 9), for "ungrouped" transient zones (defined as zones in which separate fire scenarios are postulated for all of the fixed and transient ignition sources), the licensee added 170 new scenarios to the Fire PRA model to capture fire spread to adjacent transient zones for those transient zones containing cable trays that extend across transient zone boundaries. The new scenarios were incorporated in the integrated analysis reported in the response to PRA RAI 98 (Reference 15) and updated LAR Attachment W (Reference 15). For fire zones where a hot gas layer is already postulated, the licensee stated that propagation to adjacent transient zones is not necessary because the hot gas layer scenario occurs at times shorter than propagation to adjacent transient zones and consists of damage to the full fire zone. For "grouped" transient zones (defined as zones in which all the fixed and transient ignition sources are failed completely at time of fire ignition), the licensee identified 120 new scenarios for which walkdowns showed that cable trays extend across transient zone boundaries. The licensee did not include these grouped scenarios in the Fire PRA model but did provide the results of an evaluation that showed an increase in the fire CDF by about 1.4E-06 per year.
In PRA RAI 68.01 (Reference 26), the NRC staff requested that the licensee clarify how these grouped scenarios would be included in post-transition plant change evaluations. In response to the RAI (Reference 17), the licensee explained that these scenarios were revised by either 1) crediting automatic suppression, 2) deleting them from the analysis because the fire propagation to the adjacent transient zone was already addressed by the hot gas layer analysis, or 3)
"ungrouping" and thus incorporating them into the Fire PRA model. The licensee added the new ungrouped scenarios to their PRA and reported that as a result of this re-evaluation, the increase_
in fire CDF due to the "grouped" scenarios was less than 1 E-07 per year. The NRC staff finds the licensee's transient zone method acceptable because it appropriately captures the impact of flame spread and fire propagation on fire scenario progression, and that the risk contribution from the "grouped" scenarios not explicitly included in the fire PRA model is insignificant.
Regarding the modeling of main control board (MCB) fire* scenarios noted in F&O FSS-82-01, in PRA RAI 66 (Reference 20), the NRC staff requested that the licensee provide justification for the CCDPs associated with MCB fire scenarios. In response to the RAI, and associated response to PRA RAI 66.02 (References 9 and 15), the licensee identified an error in the fault tree model logic used to evaluate failure to safely shutdown the reactor for scenarios in which main control room (MCR) abandonment is required and revised the PRA model to correct the error. The licensee also explained that the corrected model required additional detailed analysis and additional cable protection modifications to achieve acceptable risk results. Additional detail on these cable protection modifications was requested by the NRC staff in PRA RAI 66.03 (Reference 26). In response to this RAI (Reference 17), the licensee identified each of the cables for which the associated function will be protected after completion of the modifications. The licensee explained that all of these cable protection modifications are included in existing Modification ECR50784 and Modification ECR50810 included in LAR Attachment S, Table S-1. Furthermore, the licensee incorporated the additional detailed modeling and cable protection modifications in the integrated analysis reported in the response to PRA RAI 98 (Reference 15) and updated LAR Attachment W (Reference 15). The NRC staff concludes this issue is resolved because the PRA
- model logic error was corrected and the additional cable protection modifications were incorporated in the integrated analysis and updated risk results.
in PRA RAls 08 (Reference 20), 10.02 (Reference 22), 10.03 (Reference 25), 66 (Reference 20),
95 (Reference 22), and 100 (Reference 26), the NRC staff requested that the licensee provide justification for the modeling of fire scenarios in which abandonment of the MCR is credited. In response to PRA RAI 08 (Reference 8), the licensee clarified that MCR abandonment is evaluated for loss of habitability due to fires in the MCR and for loss of control due to fires in Fire Areas CB04, CB06, CB15, and CB17 (MCR). In response to PRA RAI 66 (Reference 9), the licensee provided the fault tree used to model accident scenarios with MCR evacuation followed by a failure to control the reactor remotely from the Control Room Evacuation Panel (CREP) and associated recovery actions. The fault tree logic includes basic events for 1) MCR evacuation due to the loss of habitability in the MCR, 2) MCR evacuation due to loss of control (inability to successfully shutdown from the MCR), 3) pre-mature MCR evacuation when control from the MCR is still available, 4) failure to evacuate the MCR when control from the MCR is lost, and 5) failure to successfully shutdown the reactor following MCR abandonment. In its responses to PRA RAls 10.02 (Reference 13), and PRA RAI 95 (Reference 12), the licensee further clarifies that the MCR abandonment logic considers both failure of equipment needed to successfully shutdown and operator failure to successfully shutdown per the abandonment procedure, including failures of instrumentation needed by the operator as a cue or indication in the abandonment procedure. In response to PRA RAI 10.03 (Reference 15), the licensee stated that operator failure to implement the MCR abandonment procedure is evaluated using a detailed HRA based on timed walkdowns of the abandonment procedure.
Furthermore, in response to PRA RAI 100 (Reference 18), the licensee stated that the detailed HRA of the abandonment actions, including the decision to abandon, was conducted using NUREG-1921 (Reference 51) methods. The licensee further explained that the resultant HEPs were generally based on the highest value of the methods evaluated in the HRA, that cognitive and execution timing assumptions were based on operator interviews and timed walkthroughs, and, in addition the time window available for evacuating the MCR was based on thermal-hydraulics calculations. In addition, the licensee clarified that the MCR abandonment fault tree is fully integrated into the Fire PRA model and therefore any given fire scenario where MCR abandonment is available can result in scenarios in which a) abandonment is not credited because successful shutdown from the CREP cannot be achieved, b) abandonment on loss of control is credited, and c) for the MCR only, abandonment on loss of habitability is credited. For those scenarios in which abandonment cannot be credited and control is lost, the licensee stated it uses a conditional abandonment failure probability (CAFP) of 1.0. For those scenarios in which abandonment is credited and control is lost, the licensee states the CAFP ranges between 0.02 and 1.0. For both of these categories, the licensee explains that the CAFP is essentially the CCDP or CLERP if it is assumed the conditional loss of control probability reduces the fire frequency and not CCDP or CLERP. For those scenarios in which abandonment is credited due to loss of habitability, the licensee states the CAFP ranges between 0.01 and 0.2 for CDF and between 1.0E-05 and 2.6E-03 for LERF. For this category, the licensee explains that the CAFP is essentially the CCDP or CLERP if it is assumed the conditional loss of habitability probability reduces the fire frequency and not CCDP or CLERP.
The NRC staff finds the MCR abandonment method acceptable because 1) the fault tree model includes basic events for the range of possible abandonment outcomes, and addresses failure of equipment and failure of operator actions, including failure of instrumentation relied on by the operator, 2) the licensee has developed a basis for the values assigned to the HEPs using HRA methods acceptable to the NRC and utilizing timing based on abandonment procedure walkdowns and thermal-hydraulic analyses, 3) the licensee has developed abandonment criteria that has been incorporated into plant procedures, and 4) complex scenarios (i.e., those that preclude shutting down the plant from the MCR) have an effective CCDP or CLERP of one, and less complex scenarios have a CDCP or CLERP based on an acceptable PRA evaluation of
- failure of equipment or failure of required operator actions.
In PRA RAI 02 (Reference 20), the NRC staff requested that the licensee explain how transient fires were placed to cover pinch points where CCDPs are highest for a given physical analysis unit (PAU). In response to the RAI (Reference 8), the licensee stated that transient and hot work fires are postulated in all open floor areas. The open floor area of each PAU is subdivided into one or more transient zones, which is discussed in response to PRA RAI 01 (Reference 8). The NRC staff concludes that the licensee's method for locating transient fires is acceptable because it appropriately captures all pinch points.
In PRA RAls 01 (Reference 20), 01.01 and 93 (Reference 22), and 01.02 (Reference 25), the NRC staff requested that the licensee provide further clarification and justification regarding a different method from NUREG/CR-6850 that was used for transient zone scenarios involving junction box fires and cable fires caused by welding and cutting (GFWG). In response to these RAls (References 8, 12, and 15), the licensee revised the treatment of GFWG fires to remove credit for suppression, to apply FAQ 13-0005 (Reference 86), and to utilize cable segments. This revised treatment of GFWG fires was incorporated in the integrated analysis reported in response to PRA RAI 98 (Reference 15) and updated LAR Attachment W (Reference 15).
To provide justification for apportioning GFWG fire ignition frequency according to the alternate method of using transient zone floor area, the licensee performed a sensitivity study. This sensitivity study also utilized an assumption related to cable segments which was inconsistent with FAQ 13-0005. The NRG staff notes that apportioning GFWG fire ignition frequency according to transient zone floor area is not consistent with NUREG/GR-6850, which apportions the fire ignition frequency to a physical analysis unit. Also, FAQ 13-0005 specifies that cable tray plan view, rather than the number of cable segments, is used to refine GDF.
The NRG staff finds that the licensee's approach for GFWG in the sensitivity analysis is acceptable because the number of cable tray segments is a reasonable surrogate for cable tray area since the assumption generally has less effect for high GDF transient zones which have many cable trays. The licensee performed this sensitivity analysis on all GFWG scenarios and showed that the change in fire GDF would not change the LAR conclusions, as supplemented by the updated LAR Attachment W. The sensitivity analysis also showed that the change in fire GDF for each transient fire zone was less than the self-approval guidelines. Based on the results of the licensee's sensitivity analysis and the changes made to revise the fire PRA model, the NRG staff concludes that, while the licensee's treatment of GFWG fires is inconsistent with NRG guidance, the use of this method is acceptable because it does not have a significant effect on the analysis results relative to the acceptance guidelines for transition or self-approval.
In regard to the treatment of junction box fires, the NRG staff noted that the licensee's method was non-conservative since it utilized the largest GGDP for a single cable tray or conduit entering a junction box rather than accounting for all cables entering the junction box. In response to related RAls, the licensee provided a sensitivity analysis in which the FAQ 13-0006 (Reference 87) guidance was applied to the junction box fire scenarios (Reference 15). In all cases in which the scenario GDF was greater than 1E-10 per year, the licensee's method utilizing the highest raceway GGDP resulted in a GDF greater than or equal to the NRG-accepted method in FAQ 13-0006 utilizing the total CGDP from all cables entering the junction box. The NRG staff finds the licensee's method for evaluating junction box scenarios acceptable because the entire junction box frequency apportioned to the transient zone is applied to either the GGDP of the full fire zone, the GGDP of the full transient zone, or the GGDP of the single conduit in the transient zone with the largest GGDP where the full frequency applied to the maximum GGDP is the more conservative aspect of NRG-accepted methods.
In PRA RAI 09 (Reference 20), the NRG staff noted that new information indicated that the reduction in hot short probabilities for circuits with control power transformers (GPTs) identified in NUREG/GR-6850 was too high and should be reduced. The licensee removed credit for GPTs in the integrated analysis reported in response to PRA RAI 98 (Reference 15) and updated LAR Attachment W (Reference 15) to address this change. Since credit for GPTs was removed in the integrated analysis and the updated LAR Attachment W risk results, the NRG staff concludes that this issue is resolved.
In PRA RAls 72 and 72.01 (References 20 and 22), the NRC staff requested that the licensee provide confirmation that the electrical cabinets (in which incipient detection systems will be installed and are being credited in the Fire PRA) do not contain fast acting components as defined in FAQ 08-0046 (Reference 57). In response to these RAls (References 8 and 12), the licensee explained that an independent validation, consisting of a field inspection and review of design documentation, was performed and concluded that none of the cabinets in Fire Areas CB06 (Relay Room) and CB15 (Upper Cable Spreading Room) to be monitored by incipient detection contain fast acting components as defined in FAQ 08-0046. Based on the results of this independent validation, the NRC staff concludes that the licensee has appropriately addressed fast acting components.
In PRA RAls 73 and 73.01 (References 20 and 22), the NRC staff requested justification for the incipient detection modeling assumption that 3.5 minutes is available for suppression prior to target damage, which corresponds to a fire height of 2.5 feet above the electrical cabinets in which incipient detection will be installed. In response to these RAls (References 8 and 12), the licensee revised the incipient detection analysis to assume a time to damage of 2 minutes, corresponding to a fire height of one foot above the cabinets. The licensee indicated that there are no targets within one foot of the top of the cabinets based on the actual conditions in Fire Areas CB06 and CB15 where the incipient detection systems wjll be installed. The licensee modified the modeling of incipient detection in its PRA and in the integrated analysis reported in the response to PRA RAI 98 (Reference 15) and updated LAR Attachment W (Reference 15) as described above. Based on the licensee's statement that 2 minutes for damage to targets above electrical cabinets with incipient detection is appropriate, and that this revised treatment was incorporated in the integrated analysis and the updated LAR Attachment W risk results, the NRC staff concludes that incipient detection is appropriately credited with respect to damage to targets above the cabinet.
In PRA RAI 90 (Reference 22), the NRC staff requested that the licensee provide justification for the human error: probability (HEP) value of 0.004 used to represent the failure of administrative controls to prevent the placement of transient combustibles in certain areas of the plant. In response to this RAI and PRA RAI 94 (Reference 12), the licensee stated that this method has been removed from the Fire PRA and replaced with the NRC acceptable method described in FAQ 12-0064 (Reference 88) for assignment of influence factors. Furthermore, the licensee incorporated this revised treatment for modeling transient fires in the integrated analysis reported in response to PRA RAI 98 (Reference 15) and updated LAR Attachment W (Reference 15). The NRC staff finds this acceptable because an acceptable method has been used. The licensee also stated in response to PRA RAI 90 that it may decide in the future to utilize the method removed from the Fire PRA, due to the NRC RAI; for modeling failure of administrative controls for transient combustibles with an HEP. The NRC staff concludes that the proposed license condition would allow the licensee to change the PRA in the future under its PRA maintenance and update program if the changes are consistent with the established program and methods that are acceptable to the NRC.
In PRA RAls 12 and 12.01 (References 20 and 22), the NRC staff requested that the licensee provide justification for the treatment of electrical raceway fire barrier system (ERFBS) in the FPRA. In response to these RAls (References 8 and 12), the licensee revised the treatment of ERFBS in the FPRA by including the risk contribution from fire-induced failure of one-hour rated fire barriers for fires lasting longer than one hour. Furthermore, the licensee incorporated this revised treatment of one-hour rated fire barriers in the integrated analysis reported in response to PRA RAI 98 (Reference 15) and updated LAR Attachment W (Reference 15). The revised treatment of one-hour rated fire barriers was incorporated into the PRA and the updated LAR Attachment W risk results. The NRC staff finds that this issue is resolved because the licensee is now using acceptable methods per NUREG/CR-6850.
In PRA RAls.15 and 15.01 (References 20 and 22), the NRC staff requested that the licensee provide justification for the assumption in the multi-compartment analysis (MCA) that equipment in the fire compartment that is spatially separated from the.adjacent compartment where the fire is located is not damaged until 20 minutes after the start of the fire. In response to these RAls (References 8 and 12), the licensee revised the MCA to only credit manual suppression in the fire zones for which detailed fire modeling predicts formation of a hot gas layer (HGL) having a temperature above the Kerite cable damage temperature in the exposing fire zone. The non-suppression probability (NSP) credited for these scenarios is calculated using the guidance in NUREG/CR-6850, Appendix P wherein the fire brigade response time is the time to HGL formation determined from fire modeling. The licensee also stated that for fire zones that do not have detailed fire modeling, fire suppression is not credited. The NRC staff considers the licensee's description of the revised MCA to be in accordance with the guidance in NUREG/CR-6850 and therefore acceptable. Furthermore, the licensee incorporated this revised MCA treatment in the integrated analysis reported in the response to PRA RAI 98 (Reference 15) and updated LAR Attachment W (Reference 15) to address this change. The revised MCA treatment was incorporated into the PRA and the updated LAR Attachment W risk results. The NRC staff finds that this issue is resolved because the licensee in now using acceptable methods.
In PRA RAI 83 (Reference 20), the NRC staff requested that the licensee provide justification for the use of a single heat release rate (HRR) for all pump oil spill fires regardless of the size of the spill. In response to this RAI (Reference 8), the licensee explained that the use of a single HRR value of 767 kilowatt (kW) for all pump oil spill fires regardless of oil spill size for some fire zones has no impact on the Fire PRA because each of the fire zones that used the single 767 kW are modeled as a full compartment burnout with damage at time zero and severity factor of one.
Nevertheless, in order to avoid confusion, the licensee stated that the documentation would be corrected to show possible different oil spill sizes for fire zones which are bounded by the assumption of full compartment burnout. The licensee included an action to complete updates to Fire PRA procedures to manage configuration control of NFPA 805 analysis documents in LAR Attachment S, Table S-2, Implementation Item 13. The NRC staff concludes that this action is acceptable because it will result in compliance with NFPA 805 and would be required by the proposed license condition. Based on the licensee's explanation that the use of different sizes for pump oil spills, has no impact on the Fire PRA results, the NRC staff finds that a full compartment burnout is a bounding analysis and therefore acceptable.
In PRA RAI 85.01 (Reference 22), the NRC staff requested that the licensee provide an assessment of the impact on the risk results from the propagation of parameter uncertainties. In response to this RAI (Reference 12), the licensee stated that since submittal of the LAR, the Fire PRA software had been updated to include the capability to propagate parametric uncertainties and that the Fire PRA model was updated to incorporate this capability. Furthermore, the licensee stated in the integrated analysis provided in response to PRA RAI 98 (Reference 15) that the updated uncertainty analysis utilizing this capability supports the conclusion that the estimated CDF and LERF are not significantly affected by state-of-knowledge correlations (SOKC). Furthermore, in response to PRA RAI 85.02 (Reference 17), the licensee showed that
~CDF and ~LERF are also not significantly affected by SOKC. The NRC staff concludes that this issue is resolved because SOKC has been included in the fire PRA model and the risk results were determined not to change the conclusions that the RG 1.17 4 risk guidelines are met.
In PRA RAI 13 (Reference 20), the NRC staff requested that the licensee add an implementation item in LAR Attachment S, Table S-2 to verify the validity of the reported change-in-risk following completion of proposed modifications and implementation items, and to include a plan of action to notify the NRC if the risk guidelines are exceeded. In response to this RAI (References 8) and PRA RAI 101 (Reference 17), the licensee developed an action and included it in LAR Attachment S, Table S-2, Implementation Item 22, as supplemented by the licensee's letter dated October 9, 2014 (Referencei 18), to validate/update the Fire PRA model to reflect the as-built modifications and completed implementation items, and, should the updated model not meet the RG 1.17 4 risk guidelines, actions will be taken to restore compliance with the guidelines. The licensee described these actions in the implementation item as potentially including re-analysis, additional modeling, procedure changes, or hardware changes to the plant and that the course of action will be specific to meet the RG 1.17 4 risk guidelines. The NRC staff concludes that the licensee's response to is acceptable because the licensee included an action to verify the validity of the reported change:..in-risk following completion of proposed modifications and implementation items, and included this action as an implementation item in LAR Attachment S, Table S-2, which would be required by the proposed license condition.
/
In response to PRA RAI 87 (Reference 12) the licensee stated that a confirmatory demonstration of the feasibility of credited recovery actions is included within the scope of LAR Attachment S, Table S-2, Implementation Item 15. The licensee also stated that these confirmatory demonstrations, which included field verification walk-throughs of updated station operating procedures, have been completed. In addition, the licensee stated that the HRA has also been updated to incorporate the results of the walk-throughs and updated station operating procedures. The updated HRA was discu~sed previously.
In PRA RAI 97 (Reference 23), the NRC staff requested that the licensee provide additional information about the type of reactor coolant pump (RCP) seal package to be installed under Modification ECR50799 identified in LAR Attachment S, Table S-1. In response to this RAI (Reference 13), the licensee clarified that it intends to install the Flowserve N9000 seal package,.
which is the seal package modeled in the Fire PRA. Since the seal failure model for the Flowserve N9000 seals have only been generally approved by the NRC for Combustion Engineering (CE) plants and VCSNS is a Westinghouse plant, the NRC staff requested, in PRA RAI 97.01 (Reference 26), that the licensee provide further information and justification for the seal logic model applied in the Fire PRA and associated basic event probabilities. In response to the RAI (Reference 17), the licensee stated that the RCP seal failure model and associated failure probabilities used in the Fire PRA are described in WCAP-16175-P-A, "Model for Failure of RCP Seals Given Loss of Seal Cooling in CE NSSS Plants" (Reference 89) for the three-stage RCP seal model [the NRC staff notes that this report was reviewed by the NRC and found acceptable for use at CE plants in a safety evaluation report (SER) dated February 12, 2007 (Reference 90)].
The licensee also stated that the Fire PRA seal model will be updated to incorporate credit for an additional abeyance seal, which was not included in the WCAP-16175-P-A seal model and is conservatively not currently credited in the Fire PRA model, and to reflect the as-built modification once installed. In further response to these RAls, the licensee st~ted the RCP seal modification is included in LAR Attachment S, Table S-2, Implementation Item 22, as supplemented by the licensee's letter dated October 9, 2014 (Reference 18), which states that the Fire PRA model will be updated following transition to NFPA 805 to reflect the as-built modifications, and the resulting change in risk will be verified to be either less than that estimated in the LAR, as supplemented, or less than the RG 1.17 4 acceptance guidelines. The NRC staff finds the licensee's Fire PRA RCP seal model acceptable because 1) the most appropriate basic event probabilities currently available and acceptable to the NRC staff are applied in the model and 2) the licensee developed an action and included it in LAR Attachment S, Table S-2, Implementation Item 22, which would be required by the proposed license condition, to update the Fire PRA to reflect the as-built modification and to take action to restore compliance with the RG 1.17 4 acceptance guidelines should the guidelines not be met after the Fire PRA is updated.
As a result of its review of the LAR, as supplemented, and based on the actions that would be required by the proposed license condition, the NRC staff concludes that the Fire PRA is 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 1.17 4 and are, therefore, 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 fire risk evaluations in order to gain further assurance that the methods and approaches used for the application to transition to NFPA 805 (Reference 1) were technically adequate. NFPA 805 has the following requirements that pertain to fire modeling used in support of the development of fire risk evaluations:
NFPA 805, Section 2.4.3.3, states that:
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," states that:
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," states that:
Acceptable engineering methods and numerical models shall only be used for applications to the extent th_ese 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," states that:
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," states that:
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 review of the acceptability of the fire modeling (first requirement). The results of the NRC staff's review of compliance with the remaining requirements are discussed in SE Sections 3.8.3.2 through 3.8.3.5.
3.4.2.3.1 Overview of Fire Models Used to Support the Fire Risk Evaluations Fire modeling was used to develop the zone of influence (ZOI) around ignition sources in.order to determine the thresholds at which a target would exceed the critical temperature or radiant heat flux. This approach provides a basis for the scoping or screening evaluation as part of the FRE.
The following algebraic fire models and correlations were used for this purpose:
Heskestad's Plume Temperature Correlation Point Source Radiation Model These algebraic models are described in NUREG-1805, "Fire Dynamics Tools (FDT5):
Quantitative Fire Hazard Analysis Methods for the US Nuclear Regulatory Commission Fire Protection Inspection Program" (Reference 47). The licensee used a version of Heskestad's plume temperature correlation as implemented in FIVE-Rev1, "EPRI Fire Induced Vulnerability Evaluation Methodology," Revision 1 (Reference 48), which allows the user to specify the fire location factor to account for wall and corner effects. 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," Volumes 3 and 4 (Reference 48). The V&V of the fire models that were used to support the Fire PRA is discussed in SE Section 3.8.3.2.
The CFAST computational fire model, Version 6 was used to determine the potential for the generation of a hot gas layer in selected zones. The fire PRA used these 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&V of CFAST is documented in NUREG-1824, Volume 5 (Reference 48).
The licensee also identified the use of the following empirical model that is not addressed in NUREG-1824.
Sprinkler Activation Model, described in Chapter 10 of NUREG-1805, also referred to as the DET ACT model.
V&V of the sprinkler activation model is documented in an SFPE Engineering Guide (Reference 91 ).
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 require further evaluation. The licensee stated that qualified personnel performed a plant walk-down to identify ignition sources, nearby targets, and safety-related SSCs and applied the empirical correlation screening tool 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 site-specific ignition sources that did not adversely affect the operation of credited SSCs or targets, following a fire. The licensee's screening was based on the 981h percentile fire HRR from the N_UREG/CR-6850 (References 41 - 43) methodology.
3.4.2.3.2 RAls Pertaining to Fire Modeling in Support of the VCSNS FREs In letters dated July 26, 2012 (Reference 20), and August 13, 2013 (Reference 21 ), the NRC staff requested additional information from the licensee. In letters dated October 10, 2012 (Reference 8), February 1, 2013 (Reference 9), April 1, 2013 (Reference 10), and October 14, 2013 (Reference 11 ), the licensee provided its responses to the requests.
In FM RAI 01.b (Reference 20), the NRC staff requested that the licensee explain why the reduction of the effective volume of the control room due to the presence of a large number of cabinets was not accounted for in the CFAST MCR abandonment time calculations. In its respo'nse to FM RAI 01.b (Reference 8), the licensee stated that the volume of the cabinets in the control room was slightly less than 14% of the total room volume. The licensee stated that it had performed the CFAST calculations for a sensitivity case in which the floor area was reduced by 15%, and determined that the smaller volume resulted in an abandonment time of 8.25 min, compared to 8.5 min in the base case. The licensee further stated that both cases assumed fire brigade arrival in 10 min and loss of HVAC.
The NRC staff concludes that ignoring the room volume reduction in the CFAST control room abandonment time calculations to account for the cabinets in the MCR is acceptable because the impact on control room abandonment time is not
. significant.
In FM RAI 01.c (Reference 20), the NRC staff requested that the licensee provide technical justification for the horizontal natural ventilation flow areas assumed in the MCR abandonment time calculations. In its response to FM RAI 01.c (Reference 8), the licensee stated that it had performed a sensitivity analysis in which the door gap was varied below and above the base case value of 0.02, with the fire brigade arriving after 10 minutes and loss of HVAC and that the gap below the door does not have a significant impact on the probability of abandonment or the average time to abandonment.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated that the gap below the door does not have a significant impact on the probability of abandonment or the average time to abandonment.
In FM RAI 01.d (Reference 20), the NRC staff requested that the licensee explain why the default HVAC drop-off and zero flow pressure values were used in the CFAST MCR abandonment time calculations, as opposed to plant-specific HVAC data. In its response to FM RAI 01.d (Reference 8), the licensee stated that the cases with loss of HVAC are the most conservative in terms of MCR abandonment and that since the drop-off and zero flow pressures only affect the cases with functioning HVAC, which lead to longer MCR abandonment times, the pressures do not affect the results of the analysis.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated that the default HVAC drop-off and zero flow pressure values are conservative.
In FM RAI 01.e (Reference 20), the NRC staff requested that the licensee provide technical justification for placing the fire at floor level in the CFAST MCR abandonment time calculations. In its response to FM RAI 01.e (Reference 8), the licensee stated that the habitability calculations are typically dominated by the visibility threshold parameter and the elevation of the hot gas layer, which in turn
- are sensitive to the smoke production rate. The licensee further stated that the fire base height was assumed to be at the floor of the control room, rather than 1 ft below the top of the electrical panel as recommended in NUREG/CR 6850, Supplement 1, because the former maximizes the volumetric smoke production rate in the CFAST calculations.
The NRC staff concludes that the licensee's response to the RAI is acceptable
- because the licensee demonstrated that placing the fire at floor level in the CFAST MCR abandonment time calculations is conservative.
In FM RAI 01.f (Reference 20), the NRC staff requested that the licensee provide technical justification for not considering a scenario involving fire spread to multiple cabinets in the CFAST MCR abandonment time calculations. In its response to FM RAI 01.f (Reference 8), the licensee explained that peak heat release rates up to 2.2 MW (case 5, bin 15; see Table E-4 in NUREG/CR-6850) were considered in the MCR abandonment time calculations and that at these high heat release rates, abandonment times were calculated to be relatively short (5 to 10 min) (i.e., before the peak heat release is reached and before the fire would propagate to adjacent cabinets).
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee provided adequate technical justification for not considering a scenario involving fire spread to multiple cabinets in the CFAST MCR abandonment time calculations.
In FM RAI 01.g (Reference 20), the NRC staff requested that the licensee provide technical justification for not postulating any transient fire scenarios in the CFAST MCR abandonment time calculations. In its response to FM RAI 01.g (Reference 8), the licensee stated that, based on the MCR abandonment time calculations that were performed, a minimum heat release rate of 500kW is necessary to cause abandonment and that since the 981h percentile peak heat release rate for transient fires is 317 kW, it is highly unlikely that a transient fire would result in MCR abandonment.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated that the minimum heat release rate to generate abandonment is higher than the 981h percentile peak heat release rate for transient fires which confirms that the unlikelihood of transient fires results in MCR abandonment.
In FM RAI 01.h (Reference 20), the NRC staff requested that the licensee provide the results of sensitivity studies for the MCR abandonment time calculations to address fire location, ambient temperature, etc. In its response to FM RAI 01.h (Reference 8), the licensee stated that a total of four sensitivity analyses were performed and that these studies assessed the effect of fire brigade arrival time and forced ventilation operability. The licensee further stated that additional sensitivity analyses were performed to assess the effect of reducing MCR volume to account for cabinet volume and the impact of varying leakage area. The licensee further stated that for the control room, a sensitivity analysis for ambient temperature was not performed because the room temperature is controlled to within a narrow temperature range and 1
the limiting parameter that led to abandonment was visibility.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee provided the results of the additional sensitivity studies for the MCR abandonment time calculations and these studies showed that varying cabinet volumes and door gaps do not impact the time to control room abandonment. The NRC staff also concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated that a sensitivity study for ambient temperature was not needed because the room temperature is controlled to within a narrow temperature range and because temperature is not the limiting factor that leads to abandonment.
The NRC staff found that the licensee's HGL calculations assume an opening of~
in. below every door for scenarios with closed doors and without mechanical ventilation which is consistent with the guidelines in NUREG/CR-6850, but is different from the leakage area used in the CFAST abandonment time calculations for the MCR fire scenarios with closed doors. In FM RAI 01.i (Reference 20), the NRC staff requested that the licensee explain why the leakage area assumptions in the HGL calculations were different from those in the MCR abandonment time study. In its response to FM RAI 01.i (Reference 8), the licensee stated that the
~-in. opening corresponds to a leakage area fraction of 0.003 and that a sensitivity analysis shows that varying the fractional leakage area from 0.001 to 0.03 has no significant impact on the time to abandonment or probability of abandonment. The licensee further stated that the sensitivity analyses were performed for the HGL calculations in which the equivalence ratio for the natural ventilation case was outside of the validation range and that in cases where the leakage area fraction was increased to bring the equivalence ratio into the validation range, the results were unchanged.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated that the use of a% in. opening below every door for scenarios with closed doors and without mechanical ventilation has no significant impact on the time to abandonment or probability of abandonment.
The NRC staff found that in fire zones where detailed fire modeling was performed, the licensee assumed that some fire suppression activity will be initiated before fire spreads from one transient zone to others. In FM RAI 01.k (Reference 20), the NRC staff requested that the licensee provide technical justification for the using this assumption. In its response to FM RAI 01.k (Reference 8), the licensee explained that the assumption that some fire suppression activity will be initiated before fire spreads from one transient zone to another is based, in part, on the recognition made in Chapter 14 of Supplement 1 to NUREG/CR-6850 (Reference 43) that "manual suppression is a continuous activity that can begin once the fire is detected, rather than rely primarily on fire brigade suppression efforts." In addition, the licensee stated that since the transient zones are relatively large, it is expected that a large fire will be necessary to propagate to adjacent transient zones, which suggests that detection and activities to control the fire will likely be under way.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee followed the guidance described in NUREG/CR-6850, Supplement 1, and provided a valid spatial justification that, due to the large size of transient zones, suppression will likely be initiated before fire spreads to another transient zone.
In FM RAI 01.m (Reference 20), the NRC staff requested that the licensee describe how the exact location of a target was determined for the scenarios to which it was assigned. In its response to FM RAI 01.m (Reference 8), the licensee*
explained that the cable and raceway database maintains a set of plant coordinates for each of the raceway sections and that the database does not have plant coordinates associated with conduits and cable end points, and therefore mapping of conduits and end points was done manually.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the cable and raceway database provided plant coordinates of each of the raceway sections and, for conduits and cable end points, the licensee mapped the target locations manually.
In FM RAI 01.n (Reference 20), the NRC staff requested that the licensee explain how suppression time was conservatively estimated in compartments that have numerous obstructions that could significantly delay detection and activation. In its response to FM RAI 01.n (Reference 8), the licensee explained that one of two approaches was used when suppression was credited: (1) DETACT calculations were used to confirm activation, or (2) for sprinkler heads in cable trays, it was assumed that sprinklers activate before nearby cable trays are damaged. In the first case, conservatism is achieved by assuming a longer than actual distance between the fire and the detector, while in the second case, conservatism is achieved by assuming the ZOI targets are damaged before suppression starts.
The licensee stated that this is conservative because there is no credit for suppression for the ignition source and targets within the ZOI and because the sprinkler will not be credited in protecting the nearby tray.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the approach used by the licensee provides adequate justification that detector activation times were conservatively determined by increasing the distance between the fire and the detector and by assuming the ZOI targets are damaged before suppression starts in compartments that have numerous obstructions.
In FM RAI 01.p (Reference 20), the NRC staff requested that the licensee explain how wall and corner effects were accounted for in the detailed fire modeling. In its response to FM RAI 01.p (Reference 8), the licensee explained that based on walkdowns, a fire location factor was assigned to each ignition source in the fire modeling database and the fire location factor (a value of 2 for a fire near a wall, 4 for a fire near a corner) was multiplied by the heat release rate to calculate the severity factor.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated that wall and corner effects were accounted for appropriately in that they were determined based on actual plant conditions observed during plant walkdowns.
In FM RAI 01.02 (Reference 21 ), the NRC staff requested that the licensee describe its approach to account for fire location effects in the CF AST hot gas layer
- calculations, and to explain where this approach was used. In its response to FM RAI 01.02 (Reference 11 ), the licensee explained that, in selected fire zones, additional CFAST HGL calculations were performed using the Image Method to quantify wall and corner effects on entrainment. The licensee further stated that no additional CFAST calculations were performed for fire zones in which HGL scenarios are already postulated without accounting for potential wall or corner effects and that including the re-calculated HGL scenarios in the fire PRA had minimal impact on the total risk.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee used the appropriate approach consistent with NUREG-1824, to account for wall and corner effects in the HGL analysis and that doing so, has minimal impact on the total risk.
The licensee stated that even though there is approximately 10% unknown cable (i.e., cable the licensee is unable to determine if the chemical composition is either thermosplastic or thermoset for 10% of the population) and the rest is known to be thermoset, that the whole plant is considered to have only thermoset cables. In FM RAI 01.q (Reference 20), the NRC staff requested that the licensee provide technical justification for ignoring the possibility that some of the unknown cable may be thermoplastic. In its response to FM RAI 01.q (Reference 8), the licensee explained that additional cable or raceway identification work reduced the number of unknown cables from 10% to less than 2% of the total linear feet of cable installed. The licensee further stated that since at least 98% of the cables in the plant are IEEE-383 qualified, that the flame spread rate and the HRR for thermoset cables was used in the fire modeling analyses. The licensee further stated that since at least 90% of the cables are Kerite, that the damage criteria for thermoplastic cabling was used throughout the plant.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee's data is based on a survey of 98% of the cables in the plant which demonstrates that the assumptions concerning cable flame spread, HRR characteristics, and cable damage thresholds are appropriate since they are based on plant conditions.
In FM RAI 01.t (Reference 20), the NRC staff requested that the licensee provide the following:
- i.
Technical justification for the use of a Response Time Index (RTI) of 130 (m-s)05 for the sprinkler heads in subzone 1825.01.02; ii.
Technical justification for using the FDTs to calculate activation of a sprinkler head located several feet below the ceiling; and iii.
An explanation of how smoke detection affects activation of the intermediate-level pre-action sprinkler.
In its response to FM RAI 01.t.i (Reference 8), the licensee explained that since the specific RTI values for these sprinklers are not available, a generic RTI of 130
- (m-s)05 was selected. The licensee further stated that a sensitivity analysis was performed in which the RTI was increased to 235 (m.s) 05, the highest value in the FDT spreadsheet for sprinkler activation (NUREG-1805, Chapter 10) and the sensitivity analysis showed that, although the higher RTI delays sprinkler activation, sprinklers activate soon enough to prevent spread of the fire to other transient zones.
In its response to FM RAI 01.t.ii (Reference 8), the licensee stated that the sprinklers near the ceiling, not the intermediate sprinklers, are credited for preventing propagation from the transient zone and that all the targets within the transient zone are considered damaged, but the targets outside of the transient zone are credited with sprinkler protection.
In its response to FM RAI 01.t.iii (Reference 8), the licensee explained that the sprinklers near the ceiling were used for the time to activation calculation, rather than the intermediate sprinklers and that a separate time to activation calculation for the smoke detectors was not performed. The intermediate sprinklers were not credited for preventing damage within the initiating transient zone. Since sprinklers are not credited for activating until after 20 minutes, the licensee
- 100 -
assumed that all targets within the initiating transient zone would be damaged and that sufficient smoke would be generated to allow for smoke detection and sprinkler activation for sprinklers outside the initial transient zone.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee provided appropriate technical justification for each of the items identified that support the applicable suppression calculations in that sprinklers activate soon enough to prevent the spread of fire to other transient zones, the analysis is conservative in that it considered all targets within the transient zone damaged, or the analysis is conservative because it did not credit intermediate level sprinklers within the initial transient zone, but did credit them in preventing fire spread to other transient zones after 20 minutes.
In FM RAI 01.w (Reference 20), the NRC staff requested that the licensee provide technical justification for using a zone model (CFAST) to determine whether damage will propagate outside the originating transient zone since the room dimension aspect ratios for CFAST were slightly exceeded for the L/H ratio. In its response to FM RAI 01.w (Reference.8), the licensee stated to justify its use of CF AST for the fire areas that were subjected to detailed fire modeling in support of the Fire PRA, the dimensionless parameters that describe the fire scenarios were calculated and compared against model validation parameters, as specified in NUREG-1824 and that for fire area CB04, the compartment aspect ratio and the equivalence ratio were outside of the validation range. The licensee further stated that a sensitivity case was run with CFAST, in which the room dimensions and the air flow openings into the room were adjusted to be within the validation range and that the results for the sensitivity case and the base case showed no hot gas layer temperatures above the damage criteria. The licensee further stated that since the results for both the sensitivity case and the base case were the same, the use of CFAST is considered justified. The licensee explained that a higher HGL temperature will be observed closer to the fire source and that is accounted for in the calculations because when a transient fire occurs, it is assumed that all targets in that transient zone are damaged. The licensee further stated that the CFAST results were only used to determine if targets outside the transient zone are damaged.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated an appropriate use of CFAST to determine whether damage will propagate outside the originating transient zone because it compared the parameters that describe the fire scenarios in the fire model to the model validation parameters in NUREG-1824 and found only one area to be outside of the validation range and for that area, a sensitivity study showed no hot gas layer temperatures above the damage criteria.
In FM RAI 02 (Reference 20), the NRC staff requested that the licensee provide the following:
- 101 -
- a.
A characterization of the 10% unknown cabling, specifically with regard to the critical damage threshold temperatures and critical heat flux threshold as described in NUREG/CR-6850.
- b.
An explanation of how raceways with a mixture of thermoset and thermoplastic cables were treated in terms of damage thresholds, and heat release rate and fire propagation.
- c.
A discussion of the impact on ZOI size due to increased HRR and fire propagation for thermoplastic cables.
- d.
A discussion of self-ignited cables and their impact to additional targets created.
- e.
An explanation of if and. how covered, partially covered, or holes in closed raceways affected the damage thresholds of cables used in the analysis.
- f.
An explanation of how the damage thresholds for non-cable components (i.e., pumps, valves, electrical cabinets, etc.) were determined.
- g.
Quantification of the impact of additional target damage due to the presence of thermoplastic cables on core damage frequency (CDF) and large early release frequency (LERF), as well as changes in CDF (tiCDF) and changes in LERF (~LERF) for those fire zones affected.
In its response to FM RAI 02.a and b (Reference 8), the licensee explained that additional cable/raceway identification work reduced the number of unknown cable from 10% to less than 2% of the total linear feet of cable installed and that since at least 98% of the cables in the plant are IEEE-383 qualified, the flame spread rate and the HRR for thermoset cables was used in the fire modeling analyses. The licensee further stated that since at least 90% of the cables are Kerite, the damage criteria for thermoplastic cabling was used throughout the plant and as a result, parts c and d of FM RAI 02 are not applicable.
Subsequent to the licensee's response to FM RAI 02, the licensee responded to FPE RAI 14.01. b (Reference 11 ), and in its response stated that based on the known and approximated values of qualified cable installed in the VCSNS power production areas of the plant, the non-IEEE 383-1974 cable installed in the plant comprises less than 2.5% of the total linear feet of cable installed: The NRC staff confirmed this number by reviewing the licensee's calculation which the licensee revised after the response to FM RAI 02 was submitted.
In its response to part e of FM RAI 02 (Reference 8), the licensee stated that metal covers for raceways were assumed not to have any effect on damage thresholds, which is conservative, and therefore acceptable.
In its response to part f of FM RAI 02 (Reference 8), the licensee stated that in locations with sensitive electronics, the panels are assumed damaged in the first-damage state in the progression of the fire scenario, where suppression is not credited and therefore, the damage threshold is not applicable because the panels within the particular transient zone impacted by the fire are failed immediately.
In its response to part g of FM RAI 02 (Reference 8), the licensee stated that no additional targets were identified.
3.4.2.3.3
- 102 -
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated that the cable damage thresholds have been established in accordance with NFPA 805 Section 2.5.
Conclusion for Section 3.4.2.3 Based on the licensee's description in the LAR, as supplemented, of the process for performing fire modeling in support of the fire risk evaluations, 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 the criteria in NUREG-0800, Section 19.2, Section 111.2.2.4~ 1 (Reference 40), for 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, (Reference 34) and RG 1.205,
- Section 4.3 (Reference 2) regarding the technical adequacy of the PRA used to support risk assessment to support transition to NFPA 805, subject to completion of the applicable actions as stated in the proposed license condition.
The NRC staff concludes that the PRA approach, methods and data are acceptable and therefore NFPA 805, Section 2.4.3.3 is satisfied to transition to NFPA 805. The NRC staff based this conclusion on the findings that: (1) the PRA model 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 resolution of the peer review and NRC staff review findings; and (3) the fire modeling used to support the development of the Fire PRA has been confirmed as appropriate and acceptable.
After implementation of both plant modifications and implementation items, such as procedure revisions, that would be required by the proposed license condition, the licensee must validate that the change-in risk estimates meet RG 1.17 4 risk acceptance guidelines. In response to PRA RAI 13 (Reference 8) and PRA RAI 101 (Reference 17), the licensee provided LAR Attachment S, Table S-2, Implementation Item 22 to validate/update the Fire PRA model to reflect the as-built modifications and completed implementation items, and, should the updated model not meet the RG 1.17 4 acceptance guidelines, to take actions to restore compliance with the guidelines. The licensee updated Implementation Item 22 in its letter dated October 9, 2014 (Reference 18).
Finally, based on the licensee's administrative controls to maintain the PRA models and quality, and using only qualified staff and contractors (as described in SE Section 3.8.3), the NRC staff concludes that the PRA maintenance process will ensure that the quality of the PRA is sufficient to support self-approval of future risk-informed changes to the FPP under the NFPA 805 license condition following completion of all implementation items described in the updated LAR Attachment S, Table S-2 (Reference 19).
3.4.3 Fire Risk Evaluations
- 103 -
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 th'e plant configuration. In accordance with the guidance in RG 1.205 (Reference 2), Section C,2.2.4, "Risk Evaluations,"
the licensee used a risk-informed approach to justify acceptable alternatives to comply with the NFPA 805 deterministic criteria. The NRC staff reviewed the following information during its evaluation of the fire risk evaluations: LAR Section 4.5.2, "Performance Based Approaches," LAR Attachment C, "NEI 04-02 Table 8-3 -Fire Area Transition," LAR Attachment W, "Risk Change Due to NFPA 805 Transition," and associated supplemental information.
Plant configurations that did not meet the deterministic requirements of NFPA 805, Section 4.2.3.1, were considered as variances from deterministic requirements (VFDRs). VFDRs that will be brought into deterministic compliance through plant modifications need no risk evaluation. The licensee identified in LAR Attachment C, NEI 04-02 Table 8-3, the VFDRs 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.
All of the VFDRs are either separation issues or represent noncompliance with one of the NFPA 805 Section 3.1 minimum design requirements for fire protection systems and features.
More detailed discussion about how VFDRs are identified is provided in SE Section 3.5.
In response to PRA RAI 10 (Reference 8), the licensee described, in general, how change in risk associated with VFDRs is determined in the Fire Risk Evaluations (FREs). The change in risk was calculated by determining the difference in CDF and LERF calculated using a compliant versus the post-transition plant Fire PRA. For the post-transition plant (i.e., variant plant), the VFDRs were incorporated into the PRA and contribute to CDF and LERF. In contrast, the compliant plant was modeled assuming all VFDRs were deterministically resolved. The response to PRA RAI 10 also explains that, in some cases, VFDRs are not modeled in either case because fire-induced failure of associated equipment would make insignificant contribution to fire risk.
These exceptions are explained in the individual fire area FREs.
The updated LAR Attachment W (Reference 15) provided in response to PRA RAI 98 (Reference 15) and updated LAR Attachment G (Reference 14) include further explanation of the modeling associated with determination of change in risk. The licensee explains that the deterministically compliant plant model is developed, except for MCR abandonment fire areas, by removing fire-induced failures associated with VFDRs from the post-transition plant Fire PRA.
This method for determining the compliant plant risk is acceptable to the NRC staff because it is consistent with the guidance in FAQ 08-0054 (Reference 59).
For fire areas within the control complex where MCR abandonment is credited, the licensee explains in updated LAR Attachment W, Section W.2.1, that the compliant plant risk is developed by requantifying the HEPs for the MCR abandonment recovery actions credited to reduce VFDR risk in the post-transition plant. This requantification is done such that these actions are assumed to be performed from the MCR or primary control station (PCS). Use of this approach can be seen from the post-transition and compliant plant HEP values presented in LAR Attachment G, Tables G-2 and G-3 (Reference 14) for the credited MCR abandonment actions. While the
- 104 -
licensee's method does not eliminate the risk of the credited MCR abandonment recovery actions in the compliant plant, the NRC staff concludes the licensee's method for resolving these VFDRs by converting the recovery actions to PCS actions is acceptable because attributing these credited actions to the PCS is representative of the compliant plant.
The updated LAR Attachment W, Section W.2.1 also explains that plant modifications are modeled in both the compliant and post-transition plant models. Accordingly, because the risk reduction resulting from non-VFDR related modifications is the same for both the compliant and post-transition plant models, there is no significant impact on the change in risk. Modifications that resolve VFDRs are also modeled in both the compliant and post-transition plant models and also have no significant impact on the change in risk results. VFDRs that are not resolved and therefore maintained in the plant are only modeled in the post-transition plant and result in an increase in risk.
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. The NRC staff further concludes that the results of these calculations for each fire area 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.
3.4.4 Additional Risk Presented by Recovery Actions The NRC staff reviewed LAR Attachment C, "NEI 04-02 Table B-3 -Fire Area Transition,"
LAR Attachment G, "Recovery Actions Transition," and the updated LAR Attachment W sections that were provided in response to PRA RAI 98 (Reference 15), during its evaluation of the additional risk presented by the NFPA 805 recovery actions. SE Section 3.2.5 describes the identification and evaluation of recovery actions.
The licensee used the guidance in RG 1.205, Revision 1 (Reference 2), for addressing recovery actions. This included consideration of the definition of primary control station (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. Conversely, any operator manual actions required to be performed outside the control room and not at the PCS were considered recovery actions.
The licensee provided an updated list of recovery actions in the updated LAR Attachment G transmitted as part of the licensee's RAI response in a February 25, 2014 letter (Reference 14).
Per this update, 24 unique operator actions are identified as recovery actions and are applied only to Fire Areas where MCR abandonment is credited, CB04 (CB East Lower Cable Spreading 425),
CB06 (CB Relay Room), CB15 (CB Upper Cable Spreading Room), and CB17 (CB Control Room/Support Area). The licensee explains in the response that all recovery actions listed in the updated LAR Attachment G are modeled specifically in the PRA using HRA. The licensee also explains that further recovery actions were not needed to reduce risk for transition to a RI/PB FPP.
- 105 -
The update of LAR Attachment G (Reference 14), as well as the update of LAR Attachment W (Reference 15) explains that the additional risk of recovery actions is considered to be bounded by the total change in risk. This is an acceptable approach to estimate the additional risk of recovery actions per FAQ 07-0030 (Reference 52). It should be rioted that the additional risk of recovery actions is not bounded by the total change in risk if non-VFDR related risk reduction modifications are credited only in the post-transition case in the change in risk. The licensee confirms, however, in response to PRA RAI 10.01 (Reference 12), that, when calculating the additional risk of recovery actions, credit for non-VFDR related modifications that reduce risk were modeled in both the compliant and post-transition cases.
The update of LAR Attachment W (Reference 15), reports the total additional risk of recovery actions for all fire areas to be 1.2E-06per year and 6.5E-09per year for CDF and LERF, respectively. The NRC staff finds these results acceptable because they are less than the acceptance guidelines in RG 1.174 (Reference 34).
Per the update of LAR Attachment G (Reference 14), the licensee reviewed all of the recovery actions for adverse impact. None of the recovery actions listed in LAR Attachf11ent G, Table G-1 was found to have an adverse impact on the Fire PRA. All recovery actions listed in LAR Attachment G were evaluated against feasibility criteria provided in the NEI 04-02 (Reference 5),
FAQ 07-0030, (Reference 52), and RG 1.205 (Reference 2). Per the updated LAR Attachment G, LAR Attachment S, Table S-2, Implementation Item 15, includes updates of the post-fire*
shutdown procedures and associated operator training and training procedures to incorporate the results of the recovery action feasibility evaluation and would be required by the proposed license condition.
The NRC staff concludes that the licensee's methods for determining the additional risk of recovery actions are acceptable because they are consistent with RG 1.205, Section 2.2.4.1 (Reference 2), and FAQ 07-0030 (Reference 52). The results of the total delta risk calculation is summarized in SE Section 3.4.6. As discussed in SE Sections 3.4.6 and 3.4.7, the NRC staff concludes that these results demonstrate that the total risk of transition, which also includes the additional risk of recovery actions, meets the. risk acceptance guidelines in RG 1.17 4 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.
3.4.6 Cumulative Risk and Combined Changes In LAR Attachment S, Table S..:1, the licensee identified a number of plant modifications being implemented to reduce plant risk rather than bring the plant into compliance with the deterministic requirements of NFPA 805 (e.g., ECR50577, ECR50780, ECR50799, and ECR50811). In an update of LAR Attachment W, Section W.2.1, provided as part of the response to RAI 98 (Reference 15), the licensee explains that these modifications are incorporated in both the compliant and post-transition plant fire PRA models. Accordingly, non-VFDR risk reduction modifications are not used to offset the change in fire risk. Though LAR Attachment W, Section
- 106 -
W.2.4 of the same update (Reference 15) explains that these modifications were used to decrease internal events CDF, this decrease in CDF does not change the conclusions about the comparison of change in risk to RG 1.17 4 risk acceptance guidelines. Therefore, the licensee's application to transition to a RI/PB FPP is a not a combined change request per RG 1.17 4, Revision 2, Section 1.1.
The total CDF and total LERF are estimated by adding the risk assessment results for internal, fire, and external hazard events. In response to PRA RAI 98 (Reference 15) the licensee identified a number of changes to PRA methods as discussed in this SE and provided revised estimates of total fire and internal events CDF and LERF in an update to LAR Attachment W, Table W-3. In response to PRA RAI 86.01 (Reference 15), the licensee also updated LAR Attachment W, Table W-3 to include an estimate of the seismic CDF and LERF. Based on the contribution of internal (including internal flooding), seismic, and fire events the licensee presents a total CDF of 7.0E-05 per year and total LERF of 5.0E-07 per year. The seismic CDF estimate of 1.5E-5 per year provided in response to PRA RAI 86.01 is slightly lower than the "simple average" CDF of 2.4E-5per year estimated for VCSNS in an NRC assessment attached to NRC memo "Safety/Risk Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants," dated September 2010 (Reference 92), using 2008 U.S. Geological Survey (USGS) seismic hazard estimates. However, in support of the estimate, the licensee identifies two sources of conservatism in NRC's estimate of seismic CDF. The method itself, which is based on the plant's High Confidence of Low Probability of Failure (HCLPF) capacity, has an acknowledged conservative bias, and the HCLPF capacity of 0.22g, used in NRC's estimate is underestimated and should be about 0.30g. In light of these conservatisms and the fact that a higher CDF seismic contribution would not change the conclusions about the total CDF and LERF, the NRC staff finds the estimated total CDF of 7.0E-5per year acceptable for this application.
Total CDF and LERF results are summarized in SE Table 3.4.6-1. The estimated total CDF and LERF are well below the RG 1.17 4 risk acceptance guidelines of 1 E-05per year and 1 E-06per year, respectively.
Table 3.4.6-1: CDF and LERF for VC Summer after Transition to NFPA 805(1)
Hazard Group CDF (per year)
LERF (per year)
Internal Events (including Internal 3.3E-06 1.0E-07 FloodinQ)
Fire Events 5.2E-05 2.4E-07 Seismic Events 1.5E-05 1.5E-07 Total 7.0E-05 5.0E-7 (1) Risk results provided in the response to PRA RAI 86 (Reference 12) and PRA RAI 98 (Reference 15).
The licensee also provided, as part of the response to PRA RAI 98 (Reference 15), updated t.CDF and t.LERF estimates for each fire area 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 planned modifications and administrative controls that will be implemented as part of the transition to NFPA 805, as well as recovery actions to reduce VFDR risk. The
- 107 -
change in risk values represent the final total change in risk estimates after implementing PRA model and method refinements to use NRG-accepted methods. The NRC staff finds the individual fire area risk increases acceptable because all of the fire area risk increases are less than the RG 1.17 4 risk acceptance guidelines of 1 E-05 per year for !lCDF and 1 E-06 per year for L'lLERF.
In addition, the update of LAR Attachment W, Table W-3 provided in the response to PRARAI 98 (Reference 15) reports the total !lCDF and L'lLERF for all fire areas to be 6.4E-06 per year and 1.5E-08 per year, respectively. The NRC staff finds the total risk increase acceptable because these values are within the risk acceptance guidelines of RG 1.17 4 (Reference 34). The licensee also reports in the updated LAR Attachment W, Table W-3 a decrease in internal events CDF and small increase in internal events LERF based on the risk reduction credited for non-VFDR related risk reduction modifications, however these changes, when applied to the change in fire risk, do not cause the change-in-risk associated with transition to NFPA 805 to increase to unacceptable levels. Based on the licensee's total risk and fire risk evaluation results, the NRC staff finds that the risk increase for each fire area associated with transition to NFPA 805 is within the RG 1.17 4 risk guidelines of 1 E-5/year for.D.CDF and 1 E-7per year for.D.LERF.
Subsequent to the update of the risk results provided in response to PRA RAI 98 (Reference 15),
the licensee provided a response to PRA RAI 68.01 (Reference 17), previously discussed in SE Section 3.4.2.2, in which additional ungrouped transient zone fire scenarios were added to the Fire PRA Model. While the licensee does not provide the CDF contribution from these additional scenarios in the RAI response, the CDF contribution from these and grouped transient scenarios not incorporated into the Fire PRA Model is reported to be less than 1.4E-06 per year. The NRC staff notes that even if this entire CDF contribution were attributable to !lCDF, the RG 1.17 4 risk acceptance guidelines would still be met.
Therefore, the NRC staff concludes that the risk associated with the proposed alternatives to compliance with the deterministic criteria of NFPA 805 is acceptable for the purpose of this application, in accordance with NFPA 805, 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 RAls.
The licensee used updated fire bin frequencies provided in NUREG/CR-6850, Supplement 1 (i.e.,
FAQ-08-0048). The guidance in FAQ-08-0048 (Reference 58) states that a sensitivity study must be performed using the mean of the fire frequency bins contained in Section 6 of NUREG/CR-6850 for those bins with an alpha value less than or equal to one. In response to PRA RAI 99 (References 15 and 16), the licensee provided the results of this sensitivity study showing that the fire CDF increases from 5.2E-05 per year to 8.5E-05 per year and the fire LERF increases from 2.4E-07 per year to 3.9E-07 per year, and that the fire.D.CDF increases from 6.4E-06 per year to 7.8E-06 per year and the fire.D.LERF increases from 1.5E-08 per year to 1.8E-08 per year. Based on the internal event and seismic CDFs presented in an update to LAR Attachment W, Table W-3 provided in response to PRA RAI 98, the total plant CDF after the
- 108 -
sensitivity study is 1.03E-04 per year. FAQ-08-0048 states that a search for appropriate defense-in-depth measure should be conducted when the sensitivity study results affects the decisions being made by exceeding an acceptance guideline value. In an update to the response to PRA RAI 07 (Reference 10), the licensee stated that the majority of the increase is from MCB scenarios and that these scenarios are analyzed conservatively, since unqualified cables rather than the qualified cables found mostly in the MCB are assumed in the MCB fire analysis. The licensee also identified fire protection measures that provide additional DID justifying sensitivity results not meeting acceptance guidelines. The licensee cites the control of temporary storage areas and control of hot work as measures not explicitly credited in the Fire PRA. The NRC staff finds the results acceptable because, consistent with the Supplement 1 guidance, the licensee explained that the MCB scenarios driving the risk increase are evaluated conservatively and also identified DID fire protection measures beyond those credited in the Fire PRA for areas driving the risk increase. No other key source of uncertainty requiring a sensitivity analysis was identified by the licensee or by NRC staff.
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 805, the NRC staff concludes the following:
The licensee's PRA used to perform the risk assessments in accordance with NFPA 805, Section 2.4.4 (plant change evaluations) and NFPA 805 Section 4.2.4.2 (fire risk evaluations), is of sufficient quality to support the application to transition the FPP to NFPA 805. The NRC staff concludes that the PRA approach, methods, tools and data are acceptable and therefore satisfy the associated requirement in NFPA 805, Section 2.4.3.3.
The licensee's PRA maintenance process is adequate to support self-approval of future risk informed changes to the FPP following the completion of Implementation Item 22. The proposed license condition would require that after implementation of both modifications and implementation items, the licensee verify that the resulting change in risk is either less than that estimated value in the LAR, as supplemented, or is within the RG 1'.17 4 risk acceptance guidelines.
The transition process included a detailed review of fire protection defense-in-depth and safety margin as required by NFPA 805. The NRC staff finds the licensee's evaluation of defense-in-depth and safety margin to be acceptable. The licensee's process followed the NRC endorsed guidance in NEI 04-02, Revision 2, and is consistent with the 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., llCDF and llLERF) associated with the proposed alternatives to compliance with the deterministic criteria of NFPA 805 (fire risk evaluations) were determined and are provided in accordance with the guidance in RG 1.205, Revision 1, and NFPA 805, Section 4.2.4.2. The NRC staff finds that
- 109 -
the changes in risk are all within the acceptance guidelines in RG 1.17 4, Revision 2, Sections 2.4, and NUREG-0800, Section 19.2. Therefore, these changes in risk are acceptable and satisfy the requirement in NFPA 805 Section 2.4.4.1.
The risk associated with the use of recovery actions was determined and provided iri accordance with the guidance in RG 1.205, Revision 1, and NFPA 805, Section 4.2.4. The NRC staff concludes that the additional risk associated with the NFPA 805 recovery actions is acceptable because the risk for each fire area that relies on a recovery action and the total risk of recovery actions are within the acceptance guidelines in RG 1.17 4.
The licensee did not utilize 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 1 ), Section 2.2.3, "Evaluating Performance Criteria" states that:
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 that:
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 that:
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," states that:
The purpose of this section is to define the methodology for performing a nuclear safety capability assessment. The following steps shall be performed:
- 11 b -
(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 nuclear safety equipment and cables (4)
Assessment of the ability to achieve the nuclear safety performance criteria given a fire in each fire area This SE section addresses the last topic regarding the ability of each fire area to meet the nuclear safety performance criteria of NFPA 805. SE Section 3.2.1 addresses the first three topics.
NFPA 805, Section 2.4.2.4," Fire Area Assessment," also states that:
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 nuclear safety performance criteria 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 performance-based approach to meet the nuclear safety performance criteria (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.
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 utilize deterministic methods for simplifying assumptions within the fire area.
This SE section evaluates the approach used to meet the NSPC on a fire area basis, as well as what fire protection features and systems are required to meet the nuclear safety performance criteria.
The NRC staff reviewed LAR (Reference 6) Section 4.2.4, "Fire Area Disposition," LAR Section 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 nuclear safety performance criteria of NFPA 805.
VCSNS unit one is divided into 70 fire areas, including 3 yard areas. The fire area analysis may be subdivided into zones and subzones to facilitate the analysis to achieve safe shutdown. Based on the information provided by the licensee in the LAR, as supplemented, the licensee performed
- 111 -
the nuclear safety capability assessment on a fire area basis for each of the 70 fire areas. 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.
SE Table 3.5.1 identifies the approach used (deterministic or performance based) in accordance with NFPA 805 Chapter 4 based on the information provided in, LAR Attachment C Table B-3, "Fire Area Transition".
T bl 3 5 1 VCSNS F A
a e.. '
ire rea an dC omp11ance St t ra egy s ummary Fire Area Area Description NFPA 805 Compliance Basis AB01 AB General Area, All NFPA 805 Section 4.2.4 Performance-Based Elevations (ex WPAA)
Approach CB01 CB General Area 412, 425 NFPA 805 Section 4.2.4 Performance-Based West Aooroach CB02 CB East Chase 400, 412 B NFPA 805 Section 4.2.4 Performance-Based Train Approach CB03 CB West Chase 400, 412 NFPA 805 Section 4.2.4 Performance-Based Approach CB04 CB East Lower Cable NFPA 805 Section 4.2.4 Performance-Based Spreading 425 Approach CB05 CB East Chase 400, 412 B NFPA 805 Section.4.2.4 Performance-Based Train Approach CB06' CB Relay Room NFPA 805 Section 4.2.4 Performance-Based Approach CB07 CB Plant Computer Room NFPA 805 Section: 4.2.3 - Deterministic Approach CB08 CB General Area West/ West NFPA 805 Section: 4.2.3 - Deterministic Approach Chase 436, 448 NFPA 805 Section: 4.2.4 - Performance-Based CB10 CB East Chase 436 Approach NFPA 805 Section: 4.2.4 - Performance-Based CB12 CB NE Chase 436 Approach CB14 CB Security Computer Room NFPA 805 Section: 4.2.3 - Deterministic Approach CB15 CB Upper Cable Spreading NFPA 805 Section: 4.2.4 - Performance-Based Room Approach CB Control Room/ Support NFPA 805 Section: 4.2.4 - Performance-Based CB17 Area Approach NFPA 805 Section: 4.2.4 - Performance-Based CB18 CB East Chase 463 Approach NFPA 805 Section: 4.2.4 - Performance-Based CB20 CB NE Chase 463 Approach NFPA 805 Section: 4.2.4 - Performance-Based CB22 CB HVAC Room A Approach NFPA 805 Section: 4.2.4 - Performance-Based CB23 CB HVAC Room B Approach CWPH Electric Fire Pump NFPA 805 Section: 4.2.3 - Deterministic Approach CWPH01 Room
- 112 -
Fire Area Area Description NFPA 805 Compliance Basis CWPH Diesel Fire Pump NFPA 805 Section: 4.2.3 - Deterministic Approach CWPH02 Room CWPH03 CWPH Circ Water Pump Area NFPA.805 Section: 4.2.3 - Deterministic Approach (Outdoor)
DB Underground Duct Bank NFPA 805 Section: 4.2.3 - Deterministic Approach (Conduits only)
DG01 DG Diesel Generator A, All NFPA 805 Section: 4.2.3 - Deterministic Approach Elevations DG02 DG Diesel Generator B, All NFPA 805 Section: 4.2.3 - Deterministic Approach Elevations FH01 FH General Area, All NFPA 805 Section: 4.2.4 - Performance-Based Elevations Approach IB01 I B Battery Room X 412 NFPA 805 Section: 4.2.3 - Deterministic Approach IB02 IB Battery Room A 412 NFPA 805 Section: 4.2.3 - Deterministic Approach IB03 IB Battery Charger Room A NFPA 805 Section: 4.2.3 - Deterministic Approach 412 IB04 IB Battery Charger Room B NFPA 805 Section: 4.2.3 - Deterministic Approach 412 IB05 I B Battery Charger Room A/B NFPA 805 Section: 4.2.4 - Performance-Based 412 Approach IB06 IB Battery Room B 412 NFPA 805 Section: 4.2.3 - Deterministic Aooroach IB07 IB Chilled Water Pump Rooms NFPA 805 Section: 4.2.3 - Deterministic Approach 412 IB08 IB HVAC Chiller Room C 412 NFPA 805 Section: 4.2.3 - Deterministic Approach IB09 IB HVAC Chiller Room B 412 NFPA 805 Section: 4.2.3 - Deterministic Approach IB10 IB Battery Room Ventilation A NFPA 805 Section: 4.2.4 - Performance-Based 423 Approach IB11 IB SWBP Cooling Unit Room NFPA 805 Section: 4.2.4 - Performance-Based B
Approach IB12 IB Speed Switch Room B NFPA 805 Section: 4.2.4 - Performance-Based Approach 1813 18 Speed/ XFR Switch Room NFPA 805 Section: 4.2.4 - Performance-Based "C"
Approach IB14 IB CREP Room A NFPA 805 Section: 4.2.4 - Performance-Based Approach 1815 IB CREP Room B NFPA 805 Section: 4.2.4 - Performance-Based Approach IB16 IB ESF SWGR Cooling Unit NFPA 805 Section: 4.2.3 -Deterministic Approach Room A IB17 IB ESF SWGR Cooling Unit NFPA 805 Section: 4.2.4 - Performance-Based Room B Approach IB18 IB Speed Switch Cooling Unit NFPA 805 Section: 4.2.3 - Deterministic Approach Room A
- 113 -
Fire Area Area Description NFPA 805 Compliance Basis 1819 18 Speed Switch Cooling Unit NFPA 805 Section: 4.2.3 - Deterministic Approach Room B IB20 I B 1 DA Switchgear Room NFPA 805 Section: 4.2.4 - Performance-Based Approach IB21 IB CROM Switchgear Room NFPA 805 Section: 4.2.4 - Performance-Based Approach IB22 IB 1 DB Switchgear (436)/
NFPA 805 Section: 4.2.4 - Performance-Based Ventilation Room 423 Approach IB23 IBA Chiller (412)/ SWBP NFPA 805 Section: 4.2.3 - Deterministic Approach Cooling (423)/ A Speed Switch Rooms (436)
IB24 I B Reactor Protection Panel NFPA 805 Section: 4.2.3 - Deterministic Approach Room IB25 I B General Area 412, 436/
NFPA 805 Section: 4.2.4 - Performance-Based WPAA463 Approach IB26 IB Electrical Chase 451 SE NFPA 805 Section: 4.2.3 - Deterministic Approach IB27 IB Diesel Generator B Cable NFPA 805 Section: 4.2.4 - Performance-Based Chase Approach MH02 B train of MH02 NFPA 805 Section: 4.2.3 - Deterministic Approach MH08 Manhole Yard North NFPA 805 Section: 4.2.3 - Deterministic Approach MH36 Manhole Yard TBD NFPA 805 Section: 4.2.3 - Deterministic Approach RB01 RB General Area, All NFPA 805 Section: 4.2.4 - Performance-Based Elevations Approach SWPH01 SWPH Elect Equip Room A NFPA 805 Section: 4.2.3 - Deterministic Approach SWPH02 SWPH Elect Equip Room C NFPA 805 Section: 4.2.3 - Deterministic Approach SWPH03 SWPH Elect Equip Room 8 NFPA 805 Section: 4.2.4 - Performance-Based Aooroach SWPH04 SWPH Ventilation Duct Room NFPA 805 Section: 4.2.4 - Performance-Based Aooroach SWPH05 SWPH Service Water Pump NFPA'805 Section: 4.2.4 - Performance-Based Area (436)/ Valve Pit Room Approach (425)
SWPH06 SWPH Cable Chase NFPA 805 Section: 4.2.3 - Deterministic Approach SWYD01 Electrical Switchyard NFPA 805 Section: 4.2.3 - Deterministic Approach (Outdoor)
TB01 TB General Area, All NFPA 805 Section: 4.2.4 - P,erformance-Based Elevations Aooroach TB02 TB Switchgear Room 412 NFPA 805 Section: 4.2.4 - Performance-Based Approach TB03 TB Switchgear Room 436 NFPA 805 Section: 4.2.3 - Deterministic Approach TB05 TB Switchgear Room 463 NFPA 805 Section: 4.2.3 - Deterministic Approach YD01 Refueling Wtr and Makeup Wtr NFPA 805 Section: 4.2.4 - Performance-Based Tank Area (Outdoor)
Approach
- 114 -
Fire Area Area Description NFPA 805 Compliance Basis YD02 Yard East of Plant (Outdoor)
NFPA 805 Section: 4.2.4 - Performance-Based Aooroach YD03 Station Transformer Area NFPA 805 Section: 4.2.3 - Deterministic Approach LAR Attachment C provides the results of these analyses on a fire area basis. For each fire area, the licensee documented:
The approach used in accordance with NFPA 805 (i.e., the deterministic approach in accordance with NFPA 805, Section 4.2.3, or the performance-based approach in accordance with NFPA 805, Section 4.2.4);
The SSCs required in order to meet the nuclear safety performance criteria; The fire detection and suppression systems required to meet the nuclear safety performance criteria; An evaluation of the effects of fire suppression activities on the ability to achieve the nuclear safety performance criteria; and The disposition of each variance from deterministic requirement (VFDR).
Each fire area is evaluated in accordance with NFPA 805, Chapter 4 requirements using deterministic criteria of Section 4.2.3 or the performance-based criteria of Section 4.2.4. For each performance based area, VFDRs are described along with the licensee's disposition with regard to either risk, safety margins, and defense-in-depth (for those areas complying under NFPA 805 Section 4.2.4.2) or performance based fire modeling (for those areas complying under NFPA 805 Section 4.2.4.1 ). Where recovery actions are credited for disposition of the variance from deterministic requirement, these actions are described in LAR Attachment G (see SE Section 3.2.5) and the risk associated with the action is reported on a fire area basis in LAR Attachment W.
3.5.1.1 Fire Detection & Suppression Systems Required to Meet the Nuclear Safety Performance Criteria 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 nuclear safety performance criteria. 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 address.ed 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 nuclear safety performance criteria.
- 115 -
The licensee performed a detailed analysis of fire protection features and identified the fire suppression and detection systems required to meet the nuclear safety performance criteria for each fire area. LAR Table 4-9, "Summary of NFPA 805 Required Suppression Systems" and LAR Table 4-10 "Summary of NFPA 805 Required Detection Systems" as supplemented, lists the fire areas and fire zones, and identifies if the fire suppression and detection systems installed in these areas are required to meet the deterministic criteria for nuclear safety capability assessment separation, credited in the performance-based evaluations of risk and as required by engineering evaluations and/or licensing actions.
The NRC staff reviewed LAR Tables 4-9, 4-10 and LAR Attachment C for each fire area to understand how fire detection and suppression contributes to meeting the fire protection defense-in-depth in regard to the planned transition to NFPA 805. Based upon its review, the NRC staff concludes that the treatment of this issue is acceptable because for those fire areas utilizing the deterministic approach, the installed suppression and detection systems met applicable the requirements of NFPA 805 Section 4.2.3. For those fire areas utilizing the performance-based approach, the licensee used performance-based methods to identify the fire detection and suppression systems required to meet the NFPA 805 nuclear safety performance criteria on a fire area basis.
3.5.1.2 Evaluation of Fire Suppression Effects on Nuclear Safety Performance Criteria NFPA 805 Section 4.2 requires an assessment of the effects of fire suppression activities on the ability to achieve the nuclear safety performance criteria. For each fire area in LAR Attachment C, the licensee provided a summary discussion of the analysis of the impact of manual and (where provided) automatic suppression effects on plant equipment as well as the mitigating features such as cabinet seals, cabinet design, floor drains, and fire brigade training. No adverse impacts on the ability to achieve the nuclear safety performance criteria were identified. The licensee's review did not identify any situations where fire suppression activities caused a loss of function requiring the creation of a VFDR. The licensee also explained that when water from fire suppression activities in an area drains into other areas, the other areas have been analyzed for the effects of the additional water. The licensee concluded that flooding from fire suppression activities is insignificant to the allowable level for the area, is lower than the allowable level for the area, is equal to the allowable level for the area, is controlled and maintained, or is contained due -
to curb heights, and therefore there is no impact on the nuclear safety performance criteria.
The NRC staff concludes that the licensee's evaluation of the suppression effects on the nuclear safety performance criteria is acceptable because the licensee evaluated the fire suppression effects on meeting the nuclear safety performance criteria and determined that fire suppression activities will not adversely affect achievement of the nuclear safety performance criteria.
3.5.1.3 Licensing Actions Based on the information provided in the LAR Attachment C, as supplemented, the licensee identified deviations from the deterministic licensing basis (i.e. item 2.c(18) of the current operating license) for each fire area that were previously approved by the NRC and will be transitioned with the NFPA 805 fire protection program. Where applicable, each of the deviations is summarized in LAR Attachment C on a fire area basis and described in further detail in LAR Attachment K, "Existing Licensing Action Transition". As described in the LAR, the li9ensee
- 116 -
utilized the process described in NEI 04-02 (Reference 5). 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 a determination that the basis of acceptability was still valid. The previous licensing actions will be transitioned into the NFPA 805 fire protection program as previously approved in accordance with NFPA 805 Section 2.2.7.
The LAR states that VCSNS was licensed to operate on August 6, 1982, which is after the date of January 1, 1979 in 10 CFR 50.48(b). The LAR also stated that the licensee consented to item 2.c(18) of the operating license which required the licensee to maintain the plant fire protection program in accordance with Section 111.G., 111.L, and 111.0., of Appendix R to 10 CFR Part 50.
Because VCSNS compliance with Appendix R is not required by 10 CFR 50.48, deviations from Appendix R provisions imposed by the license condition were not approved via exemptions, but were evaluated in NRC safety evaluations. Since the previously approved deviations are either compliant with 10 CFR 50.48(c) or no longer necessary, as discussed in LAR Attachment M, upon issuance of the new 10 CFR 50.48(c) license condition, the current VCSNS license condition will be superseded. The licensee understands that implicit in the superseding of the current license condition, all prior fire protection program safety evaluation reports and commitments will be superseded in their entirety.
The licensee does not have any elements of the current fire protection program for which NRC clarification is needed. The licensing actions being transitioned are summarized in SE Table 3.5-2.
- 117 -
Table 3 5-2 Previously Approved Licens1nq Actions Beinq Transitioned Licensing Action Description Applicable Fire Areas LA-AB01-01 -Appendix R AB01 Deviation, Auxiliary Building -
Lack of 20 ft of separation and automatic suppression (111.G.2.b criteria)
LA-AB01-02 -Appendix R AB-01 Deviation, Auxiliary Building -
Lack of automatic suppression (111.G.2 criteria)
NRC Staff Evaluation
The basis for approval as described by the licensee in LAR Attachment K is that Train B cable in Fire Zone AB01.09 is separated from Train A cable in Zones AB01.10, AB01.18, and AB01.21 by one to three 3-hour rated barriers (floors) with unprotected openings, Cable trays are provided with fire stops where they penetrate the floor, Automatic detection in each affected fire zone, and fire suppression is provided by interior manual hose stations and portable extinguishers.
Based on the previous staff approval of this deviation in an SE dated 7127187 (Reference 32) and inclusion in the existing FP license condition, and the statement by the licensee that the basis remains valid, the NRC staff concludes that transition of this licensing action is acceptable.
The basis for approval as described by the licensee in LAR Attachment C, is that the existing deviation is "Adequate for the Hazard" based on the following conditions:
Fire zone AB01.09 is separated from adjacent fire zones by fire rated construction with limited unsealed openings Redundant charging pump circuits wrapped throughout fire zone Greater than 20 foot of separation between ttie "A" and "B" Charging Pump Cooling Units Redundant cooling unit circuits wrapped throughout zone providing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire barrier Automatic detection throughout fire zone Partial automatic suppression
- 118 -
Licensing Action Description Applicable Fire
NRC Staff Evaluation
Areas throughout fire zone Modification to expand automatic suppression to include entire fire zone (ECR 50810)
Modification to wrap unprotected miniflow valve circuits in fire zone (ECR 50784)
- Manual suppression capability with extinguishers Based on the previous staff approval of this deviation in an SER dated 8/20/82 (Reference 29) and inclusion in the existing FP license condition, and the statement by the licensee that the basis remains valid, the NRC staff concludes that transition of this licensing action is acceptable.
LA-CB02-'01 -Appendix R CB02 The basis for approval as described by Deviation, Control Building -
the licensee in LAR Attachment K is that Lack of 1-hour fire rated barrier 1-hr rated Rockbestos Firezone R fire (111.G.2.c criteria) resistant cables were used in lieu of a 1-hr wrap.
Based on the previous staff approval of this deviation in an SE dated 10/19/97 (Reference 99), and the statement by the licensee that the basis remains valid, the NRC staff concludes that transition of this licensing action is acceptable.
LA-1607 Appendix R IB07 The basis for approval as described by Deviation, Intermediate the licensee in LAR Attachment K is that Building - Lack of 20-ft the area has an automatic sprinkler separation (1111.G.2.b criteria) system and a fire detection system installed. 1-hr rated radiant shield walls between all three pumps are provided to divide the room into three areas.
Additionally, a 1-hr rated fire barrier has been provided for a cable from one division which passes through the pump area for another division.
Based on the previous staff approval of this deviation in an SER dated 1/8/82
- 119 -
Licensing Action Description Applicable Fire
NRC Staff Evaluation
Areas (Reference 28), and inclusion in the existing FP license condition, and the statement by the licensee that the basis remains valid, the NRC staff concludes that transition of this licensing action is acceptable.
LA-1825-01 -Appendix R 1825 The basis for approval as described by Deviation, Intermediate the licensee in LAR Attachment K is Building - Lack of 20 foot smoke detection system installed, separation (111.G.2.b criteria) sprinkler system to cover CC pumps and extend at least 15-ft beyond each pump (subsequently, full automatic suppression was installed throughout the area), 1-hr f,ire rated barrier on one division if redundant separation is less than 20-ft of clear space (no combustibles), 10-ft high radiant heat shield wall constructed of drywall between pumps Band C (only one CC pump required).
Based on the previous staff approval of this deviation in an SER dated 1/8/82 (Reference 28),.and inclusion in the existing FP license condition, and the statement by the licensee that the basis remains valid, the NRC staff concludes that transition of this licensing action is acceptable.
LA-IB25 Appendix R 1825 The basis for approval as described by Deviation, Intermediate the licensee in LAR Attachment K is Building - Lack of 1-hour fire redundant circuits are separated rated barrier (111.G.2.c criteria) horizontally by 12-ft and by a reinforced concrete wall with unprotected openings. Automatic suppression and detection is installed in fire zone 1825.01 and automatic detection is installed in the Train B cable chase. Also there is a 3-hr fire barrier with unprotected openings around Train B cable chase.
Based on the previous staff approval of this deviation in an SE dated 7127187 (Reference 32), and inclusion in the existing FP license condition, and the
- 120 -
Licensing Action Description Applicable Fire
NRC Staff Evaluation
Areas statement by the licensee that the basis remains valid, the NRC staff concludes that transition of this licensing action is acceptable.
LA-IB25 Appendix R IB25 The basis for approval as described by Deviation, Intermediate the licensee in LAR Attachment K is that Building - Radiant energy there are pre-action sprinklers above shield in lieu of a 1-hour fire and below the B&W Kaowool M-board.
rated barrier (1111.G.2.c criteria) 1 /2" diameter hanger rods enclosed with 1 /2" wall thickness of Thermo-Lag 330-1 split tubing is credited as equivalent to 1-hr fire rated barrier. Also coated surfaces of Unitstrut with TSI material (trowel grade or flexible wrap) is credited as equivalent to a 1-hr fire rated barrier. Fusible-type water spray nozzles are provided for cable tray stacks in the overhead and the fire area is protected by automatic fire detection and suppression.
The top part of "M" board is covered by 1 /16" thick fire-retardant *"Tuff Span" sheeting to provide mechanical damage protection and the pipe penetrations are sealed with kaowool blankets.
Based on the previous staff approval of this deviation in an SE dated 5/22/86 (Reference 30), and inclusion in the existing FP license condition, and the statement by the licensee that the basis remains valid, the NRC staff concludes that transition of this licensing action is acceptable.
LA-IB25-04-Appendix R IB25 The basis for approval as described by Deviation, Lack of 3-hour the licensee in LAR Attachment K is as barrier (111.G.2.a criteria) follows:
Either Channel A or Channel B Core exit thermocouples (T/C) will also be available in the four fire zones (2 per quadrant).
- Alternate methods to determine the
- 121 -
Licensing Action Description Applicable Fire
NRC Staff Evaluation
Areas existence of natural circulation cooling.
- Direct Method - Utilize SG pressure as a substitute for Tcold.
- Indirect Method - Use RCS temperature (Thot), RCS pressure, and steam tables to assure RCS is subcooled and water solid.
Based on the previous staff approval of this deviation in an SE dated 11/26/86 (Reference 31) and inclusion in the existing FP license condition, and the statement by the licensee that the basis remains valid, the NRC staff concludes that transition of this licensing action is acceptable.
LA-IB25-05 -Appendix R IB25 The basis for approval as described by Deviation, Various Areas in the the licensee in LAR Attachment C and K Auxiliary and Intermediate is due to the existing deviation for a lack Buildings Lack of Automatic of full automatic fire suppression Suppression (111.G.2 criteria) throughout fire zone IB25.06 as demonstrated by an "Adequate for the Hazard" Fire Protection Engineering Equivalency Evaluation based on the NSCA.
Also, automatic smoke detection is provided throughout fire zone IB25.06.
Based on the previous staff approval of this deviation in an SER dated 8/20/82 (Reference 29), and inclusion in the existing FP license condition, and the statement by the licensee that the basis remains valid, the NRC staff concludes that transition of this licensing action is acceptable.
LA-1825 Appendix R 1825 The basis for approval as described by Deviation, Intermediate the licensee in LAR Attachment K is that Building - Lack of 1-hour fire 1-hr cables are installed in lieu of rated barrier (111.G.2.c criteria) enclosing Train A tray 3088 in 1-hour fire wrap throughout FA IB-25.
Based on the previous staff approval of
- 122 -
Licensing Action Description Applicable Fire
NRC Staff Evaluation
Areas this deviation in an SE dated 10/19/97 (Reference 99), and the statement by the licensee that the basis remains valid, the NRC staff concludes that transition of this licensing action is acceptable.
LA-MH02-01 -Appendix R MH02 The basis for approval as described by Deviation, Man Hole - Lack of the licensee in LAR Attachment K is that 3-hour fire rated barrier MH-2.1 and MH-2.2 are separated by a (111.G.2.a criteria) Redundant 6" concrete wall with a 4" pipe opening trains for SW Pump House are at the base for drainage. There is a not separated by a fire barrier 2-ft thick concrete manhole cover that having 3-hour fire rating.
precludes entry of transient MH-2.1 - contains A train, combustibles. There is also low MH-2.2 -contains B train.
combustible loading consisting only of cable insulation.
Based on the previous staff approval of this deviation in an SE dated 7/27/87 (Reference 32), the inclusion in the existing FP license condition, and the statement by the licensee that the basis remains valid, the NRC staff concludes that transition of this licensing action is acceptable.
LA-RB01 Appendix R RB01 The basis for approval as described by Deviation, Intermediate the licensee in LAR Attachment K is that Building - Lack of 3-hour fire either Channel A or Channel B Core exit rated barrier (111.G.2.a criteria) thermocouples (TIC) will also be Redundant power for Th and available in the four fire zones (2 per Tc not separated by 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
quadrant). Alternate methods to determine the existence of natural circulation cooling are the Direct Method which uses SG pressure as a substitute for Tcold and the Indirect Method which uses RCS temperature (Thot), RCS pressure, and steam tables to assure RCS is subcooled and water solid.
Based on the previous staff approval of this deviation in an SE dated 11/26/86 (Reference 31 ), the inclusion in the existing FP licensing basis, and the statement by the licensee that the basis remains valid, the NRC staff concludes
- 123 -
Licensing Action Description Applicable Fire
NRC Staff Evaluation
Areas that transition of this licensing action is acceptable.
LA-FEAT-04 -Appendix R Various The basis for approval as described by Deviation, Intermediate the licensee in LAR Attachment K is that Building - Lack of 3-hour fire bullet resistant and pressure doors, rated door (111.G.2.a criteria) doors that were manufactured of similar materials and construction to rated fire doors and doors that do not have any openings or ports, and are self-closing, were found to be acceptable in the areas where they were used.
Based on the previous staff approval of this deviation in an SER dated 1/8/82 (Reference 28), the inclusion in the existing FP license condition, and the statement by the licensee that the basis remains valid, the NRC staff concludes that transition of this licensing action is acceptable.
LA-FEA T Appendix R Various The basis for approval as described by Deviation, Intermediate the licensee in LAR Attachment K is that Building - Lack of 3-hour fire the fire dampers are dual 1.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> rated rated damper (111.G.2.a criteria).
fire dampers in lieu of a 3-hour rated Back-to-back dual 1.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> dampers, automatic detection is rated fire dampers in lieu of a installed in areas where these dampers 3-hour rated fire dampers are located and the areas have a low fire loading or automatic detection and suppression is installed in areas where high fire loading exists.
Based on the previous staff approval of this deviation in an SER dated 1/8/82 (Reference 28), the inclusion in the existing FP license condition, and the statement by the licensee that the basis remains valid, the NRC staff concludes that transition of this licensing action is acceptable.
On the basis of: (1) The process described in LAR Section 4.2.3, "Licensing Actions: Resulting NFPA 805Analysis & Transitions" and utilized by the licensee to evaluate the licensing actions; (2)
The licensee's determination that the basis of acceptability remains valid; and (3) The NRC staff's review of each licensing action, the basis for previous NRC staff approval, and the applicability to the new requirements, the NRC staff concludes that the transition of the licensing actions
- 124 -
described in LAR Table 4-3 "NSCA FPEEs/Licensing Actions," LAR Attachment C and LAR Attachment K is acceptable because the applicability to the new requirements and the original basis of acceptability remains valid.
3.5.1.4 Existing Fire Protection Engineering Equivalency Evaluations (FPEEEs)
In LAR Section 4.2.2, the licensee stated that the fire protection engineering equivalency evaluations that support compliance with NFPA 805 Chapter 3 or Chapter 4 were reviewed using the methodology contained in NEI 04-02 (Reference 5) and that the methodology for performing the FPEEE review included the following determinations:
The FPEEE is not based solely on quantitative risk evaluations; The FPEEE is an appropriate use of an engineering equivalency evaluation; The FPEEE is of appropriate quality; The standard license condition is met; The FPEEE is technically adequate; The FPEEE reflects the plant as-built condition; and The basis for acceptability of the FPEEE remains valid.
In LAR Section 4.2.2, the licensee stated that in accordance with the guidance in RG 1.205 (Reference 2), Regulatory Position 2.3.2, as clarified by FAQ 08-0054, (Reference 59) FPEEEs 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" FPEEEs, then the FPEEE was referenced and a brief description of the evaluated condition was provided. These are referenced in the Attachments A and C as appropriate.
If requesting specific NRC approval for "adequate for the hazard" FPEEEs, then FPEEE was referenced to demonstrate compliance and was included in Attachment Kor Attachment L, as appropriate for NRC review and approval.
The licensee identified and sqmmarized the FPEEEs for each fire area in LAR Attachments A and C, as appropriate. The licensee requested NRC review and approval of those engineering evaluations listed in LAR, Section 4.1.2.3, "NFPA 805 Chapter 3: Compliance Alternatives (CA)
Not Previously Approved by NRC" and LAR Table 4-1, "NFPA 805 Chapter 3 Requests for Approval." These engineering evaluations are discussed in SE Section 3.1.4.
Based on the NRC staff's review of the licensee's methodology for review of FPEEE's and identification of the applicable FPEEEs in LAR Attachment C, the NRC staff concludes that the use of FPEEEs is acceptable because they meet the guidance provided in RG 1.205 and FAQ 08-0054, and the requirements of NFPA 805.
- 125 -
3.5.1.5 Variances from Deterministic Requirements (VFDRs)
For those fire areas where deterministic criteria were not met, variances from deterministic requirements were identified and evaluated using performance-based methods. Variances from deterministic requirements 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 variances from deterministic requirements:
A fire risk evaluation determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluations; A fire risk evaluation determined that the risk, safety margin a.nd defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluations and crediting existing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rated ERFBS; A fire risk evaluation determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluations and an engineering change request (ECR) package; A fire risk evaluation determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a fire risk evaluation and administrative control to preclude equipment failure; A fire risk evaluation determined that the risk, safety margin and defense-in-.depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluations and installing electrical raceway fire barrier systems; A fire risk evaluation determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluations and recovery actions;
.A fire modeling performance-based evaluation determined that the margin between the maximum expected fire scenario and the limiting fire scenario demonstrated that one success path would remain free of fire damage; A fire modeling performance-based evaluation determined that the margin between the maximum expected fire scenario and the limiting fire scenario demonstrated that one success path would remain free of fire damage considering existing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rated electrical raceway fire barrier systems; and A fire risk evaluation determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with alternate shutdown recovery actions.
For all fire areas where the licensee utilized the performance-based approach to meet the nuclear safety performance criteria, each variance from deterministic requirement and the associated disposition has been described in LAR Attachment C. For those fire areas using the fire modeling
- 126 -
performance-based approach, the licensee determined that the margin between the maximum expected fire scenario and the limiting fire scenario demonstrated that one success path would remain free of fire damage. For those firea areas using the fire risk evaluation performance-based approach, an integrated assessment of risk, defense-in-depth and safety margins concluded that the conditions were acceptable. Based on the review of the variances from deterministic requirements and associated resolutions as described in LAR Attachment C, the NRC staff concludes that the licensee's identification and resolution of the variances from deterministic requirements is acceptable because the performance-based evaluations demonstrate that the nuclear safety performance criteria are met.
3.5.1.6 Recovery Actions LAR Attachment G, Table G-1, lists the recovery actions identified in the resolution of variances from deterministic requirements in LAR Attachment C for each fire area. The* recovery actions identified were actions considered necessary to meet risk acceptance criteria.
The NRC staff reviewed LAR Section 4.2.1.3, "Establishing Recovery Actions, and LAR Attachment G, "Recovery Actions Transition, to evaluate whether the licensee meets the associated requirements for the use of recovery actions per NFPA 805. The details of the NRC staff review for recovery actions are described in SE Section 3.2.4 "Establishing Recovery Actions."
3.5.1.7 Recovery Actions Credited for Defense in Depth (RA-DID)
VCSNS does not utilize recovery actions credited for defense-in-depth.
3.5.1.8 Plant Fire Barriers and Separations Passive fire protection features (e.g., fire barriers, through penetration fire stops, and penetration seqls) and active fire protection features (e.g., doors, dampers, and water curtains) include the fire barriers and the associated elements used to form fire area boundaries and barriers separating success paths necessary to meet the nuclear. safety performance criteria. The fire barrier fire resistance rating necessary for separation between fire areas under NFPA 805 (i.e., 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) is the same as that necessary under the plant's pre-NFPA 805 licensing basis. Where the fire barriers do not meet the required fire resistance rating, the licensee has performed FPEEEs on the acceptability of the barrier relative to the hazards in the fire area, as discussed in SE Section 3.5.1.4.
In addition to those established fire barriers and separations that define the plant fire areas, passive fire protection features may include such design elements or features as radiant energy shields, flame impingement shields, high-energy arching fault (HEAF) shields and electrical raceway fire barrier systems that are credited with protecting cables, electrical components, and equipment within a fire area from the effects of fire or high-energy faults.
LAR Table 4-3, "NSCA FPEEs/Licensing Actions," as supplemented, documents the FPEEEs, the licensing action and a brief description. LAR Attachment K "Existing Licensing Action Transition" lists the Fire Area, Licensing Action Number, whether it is to transition to NFPA 805 and pertinent NFPA 805 comments. Additionally, the type of deviation (i.e., lack of automatic
- 127 -
suppression, lack of 1-hour fire rated barrier), detailed technical information, and a basis are provided for each licensing action. Similar information to LAR Attachment K can also be found in LAR Attachment C, "NEI 04-02 Table B-3 Fire Area Transition" on a fire area basis. These documents identify passive fire protection features that are required to meet NFPA 805 criteria.
Fire Area DB, underground duct bank, credits conduits within duct banks that are encased in at least 5" of concrete, 6" of concrete that separate the channels of redundant circuits, and single channel duct banks are isolated from one another by 30'. Embedded conduits are also credited in fire area DB. Fire area IB07, IB Chilled Water Pump Rooms 412, credits 3-hour fire barriers,
between the chilled water pump rooms from adjacent areas and existing 1-hour rated radiant energy shields. Fire area MH02, B Train of MH02, in the manhole, credits 6" concrete wall (containing drainage opening at bottom of wall) between redundant circuits with a 2' thick concrete cover.
SE Section 3.1 provides the results of the NRC staff's evaluation of the acceptability of fire barriers and separations against the NFPA 805 Chapter 3 Section 3.11 minimum design requirements for these fire protection features.
3.5.1.9 Electrical Raceway Fire Barrier Systems (ERFBS)
The licensee used ERFBS in both 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rated configurations. An example of a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rated ERFBS is documented in the licensee's analysis for fire area IB07, IB chilled water pump room 412, credits existing 1-hr rated electrical raceway fire barrier systems and meets the deterministic criteria of NFPA 805, Section 4.2.3.3(c). An example of a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rated ERFBS is documented in the licensee's analysis for fire area IB23, 1 BA chiller (412)/~WBP cooling (423)/A speed switch rooms (436), credits existing 3-hr rated electrical raceway fire barrier systems and meets the deterministic criteria of NFPA 805 Section 4.2.3.3(a). The LAR states that required electrical raceway fire barrier systems are installed to provide a 1-hour or 3-hour fire barrier rating and that qualification testing has been performed in accordance with Generic Letter 86-10, Supplement 1 (Reference 93) or equivalent performance testing to ensure the protected raceways are free of fire damage.
The results of the NRC staff's evaluation of the acceptability of the electrical raceway fire barrier systems against the NFPA 805 Chapter 3 Element 3.11.5 minimum design requirements are provided in SE Section 3. 1.
3.5.1.10 Issue Resolution The licensee plans to change to a reactor coolant pump seal package specifically designed to provide low leakage under shutdown conditions so that the loss of seal cooling does not lead to significant loss of reactor coolant system inventory. Based on information provided in the licensee's December 11, 2014 letter (Reference 19), the new RCP seals are currently planned to be installed by December 31, 2015. The licensee stated in the LAR that until the new seal materials are installed, procedures for seal cooling interruptions are in place to address the issue as part of the existing Appendix R analysis. In its response to SSD RAI 12 (Reference 8), the licensee further stated that, since the installation of these seals is expected to extend beyond implementation of the NFPA 805 program, these procedures would continue to be maintained under the approved NFPA 805 program until the planned seal modifications are complete.
- 128 -
The licensee's current method to meet the technical requirements of 10 CFR 50, Appendix R, Section 111.G.3 involves the use ofSelf-lnduced Station Blackout (SISBO). This methodology involves intentionally de-energizing both offsite power and one on-site emergency power source to prevent spurious operation of equipment. Also, local operator manual actions are credit~d to position or verify position of motor and pneumatic operated valves. The SISBO methodology will be eliminated during the NFPA 805 transition process. The licensee included actions in LAR Attachment S, Table S-2, Implementation Items 15 and 16, which state that the licensee will update station operating procedures (which are not modification related) and conduct associated training to incorporate insights and the changes in operational shutdown strategy in response to a fire, and complete administrative procec;jures and documents to support the implementation of the non-power modes of plant operating states for NFPA 805. The NRC staff concludes that these actions are acceptable because they will result in compliance with NFPA 805 and would be required by the proposed license condition.
The elimination of the SISBO methodology will include analysis (determining the extent of fire damage to safe shutdown equipment and cables in each fire area, performing the detailed cable identification, circuit analysis, and physically locating cables that are required for the safe shutdown equipment}, and procedure development, implementation and training. Procedure development will utilize insights obtained from the risk-informed, performance-based analyses performed to meet NFPA 805 Chapter 2.
3.5.1.11 Conclusion for Section 3.5.1 As documented in LAR Attachment C, as supplemented, 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 meets the associated criteria of NFPA 805, Section 4.2.3. This conclusion is based on:
The licensee's documented compliance with NFPA 805, Section 4.2.3; Deviations from the pre-NFPA 805 fire protection licensing basis were reviewed by the licensee for applicability, as well as continued validity, and found acceptable; The licensee's assertion that the success path will be free of fire damage without reliance on recovery actions; The licensee's assessment that the suppression systems in the fire area will have no impact on the ability to meet the nuclear safety performance criteria; and The licensee's appropriate determination of the automatic fire suppression and detection systems required to meet the nuclear safety performance criteria.
For those fire areas that utilized the performance-based approach in accordance with NFPA 805, Section 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:
- 129 -
Deviations from the pre-NFPA 805 fire protection licensing basis that were transitioned to the NFPA 805 licensing basis were reviewed by the licensee for applicability, as well as continued validity, and found acceptable; For those fire areas using the fire risk evaluation performance based approach in accordance with NFPA 805, Section 4.2.4.2, VFDRs were evaluated by the licensee and found acceptable based an integrated assessment of risk, defense-in-depth, and safety margins. Where credited in the disposition of the variances from deterministic requirements, modifications and recovery actions were identified and actions planned to address the issue (see SE Sections 3.4.1 and 3.4.6);
For those fire areas using the fire modeling performance based approach in accordance with NFPA 805, Section 4.2.4.1, VFDRs were evaluated by the licensee and found acceptable using performance based fire modeling.
Implementation items address the modifications and other actions as applicable.
Recovery actions used to demonstrate the availability of a success path to achieve the nuclear safety performance criteria were evaluated by the licensee and the additional risk of their use;*determined, reported, and found to be acceptable (see SE Sections 3.4.1 and 3.4.6);.
The licensee's analysis appropriately identified the fire protection systems, structures and components required to meet the nuclear safety performance criteria, including fire suppression and detection systems, as well as required fire protection features; Fire area boundaries (ceilings, walls, and floors), such as fire barriers, fire barrier penetrations, and through penetration fire stops have been established by the licensee and the NRC staff finds them acceptable; Electrical raceway fire barrier systems that are credited in meeting the requirements of NFPA 805 are documented on a fire area basis and verified to meet the criteria of NFPA 805; and The fire risk meets the NFPA 805 acceptance criteria of being acceptable to the AHJ (NRC) as required by NFPA 805, Section 2.4.4.1, and the philosophy of defense-in-depth and safety margin are maintained.
Accordingly, the NRC staff concludes that each fire area utilizing the deterministic or performance-based approach meets the applicable requirements of NFPA 805, Section 4.2.
3.5.2 Clarification of Prior NRC Approvals As stated in LAR Attachment T, there are no elements of the pre-transition licensing basis that require clarification of prior NRC approval.
- 130 -
3.5.3 Fire Protection During Non-Power Operational 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 LAR Attachment D, "NEI 04-02 Non-Power Operational Modes Transition," to evaluate the licensee's treatment of potential fire impacts during non-power operations (NPOs). The NRC staff concludes that the licensee followed the guidance used in the process described in NEI 04-02, (Reference 5) as modified by FAQ 07-0040 (Reference 56), for demonstrating that the nuclear safety performance criteria are met for higher risk evolutions (HREs) during NPO modes.
3.5.3.1 NPO Strategy and PlantOperational States (POSs)
LAR Section 4.3, "Non-Power Operational Modes," and LAR Attachment D, "NEI 04 Non-Power Operations Modes Transition," describe the licensee's implementation of the FAQ 07-0040 (Reference 56) process. The licensee stated that its goal is to ensure that contingency plans are established when the plant is in an NPO condition where the risk is high. The licensee's strategy for control and protection of equipment during NPO modes is shown in LAR Figure 4-6, which considers the availability of Key Safety Function (KSF) equipment, if the KSF may be lost due to fire, and if the plant is in a HRE, the use of contingency plans to mitigate the risks. If the plant is not in a HRE, normal risk management controls and fire prevention/protection processes and procedures are utilized.
As described in LAR Attachment D, the licensee's procedure defines HREs as "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 procedure contains specific actions to address reduced inventory conditions that result in a short time to boil, limited methods for decay heat removal, and low RCS inventory and are consistent with those described in FAQ 07-0040.
The NRC staff concludes that the NPO process dJscribed and documented by the licensee in LAR Section 4.3 and LAR Attachment Dis consistent with NEI 04-02 and FAQ 07-0040 and is acceptable because the licensee has (1) identified: the KSFs required to maintain the plant in a safe and stable condition during NPO, (2) identified those conditions that constitute HREs, and, (3) established contingency plans to mitigate fire risk during HREs.
- 131 -
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 Jose 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 process~s and procedures will be utilized.
Consistent with the guidance in NEI 04-02 (Reference 5) and FAQ 07-0040 (Reference 56), the process to demonstrate that the nuclear safety performance criteria are met during NPO modes I
involved the following steps as described in LAR Sections 4.3.1 and 4.3.2, and depicted in LAR Figures 4-5 and 4-6:
Review of the existing Outage Manag~ment Processes.
Identification of Equipment/Cables:
i
- Review of plant systems to determine success paths that support each of the defense-in-depth KSFs;. arid I
Identification of cables required for the selected components and determination of their routing.
Perform Fire Area Assessments (identify pinch points - plant locations where a single fire may damage all success paths of a KSF).
Manage pinch-points associated with fire-induced vulnerabilities during the outage.
3.5.3.3 NPO Pinch Point Resolutions and Program Implementation The licensee performed reviews of each fire area where equipment or cabling associated with KSF success paths might be damaged as a result of a fire. These reviews identified that there are fire areas where a single fire could result in a loss of all credited paths for a given KSF (i.e. pinch point). The review also identified that there are certain fire areas that are vulnerable to a loss of a KSF if certain system trains or components are taken out of service during a non-power operational mode and a fire were to occur. Identified pinch points were resolved through a combination of procedural changes, crediting modifications that are to be implemented as a result of the revali_dated post-fire safe shutdown analysis, or use of recovery actions.
Assessments performed by the licensee as part of the NPO review conservatively assumed that all NPO components or component cables in the fire area may be lost due to a fire. Use of the review methodology outlined in the LAR and the approaches developed to alleviate the identified "pinch points" precluded the need to utilize fire modeling in order to resolve a KSF concern.
Approximately 50 pinch points were identified during the performance of the NPO fire area reviews, involving 41 individual components.
- 132 -
The LAR states that the identified pinch points and the fire areas where a fire might cause
- damage to equipment required to support a KSF path are documented in a site-specific calculation. The results of the non-power modes transition review, in accordance with the methodology described in NFPA 805 (Reference 1) Section 2.4.2 and NFPA 805 Appendix B.2, NEI 04-02 (Reference 5) Section 4.3.3 and NEI 04-02 Appendix F, and FAQ 07-0040 (Reference 56), are also documented in this calculation.
During those NPO evolutions where risk is relatively low, the normal fire protection program defense-in-depth actions are credited for addressing the risk impact of those fires that potentially impact one or more trains of equipment that provide a KSF required during non-power operations.
During these relatively low risk scenarios, control of ignition sources, control of combustibles and compensatory actions for fire protection system impairments are considered to be adequate to address minor losses of system capability or redundancy.
During the NPO evolutions that are defined as HREs, additional fire protection defense-in-depth measures will be taken. These are managing risk in fire areas that contain known pinch points and managing risk in fire areas where pinch points may arise because of equipment taken out of service.
In LAR Attachment S, Table S-2, Implementation Item 16 identifies an action to incorporate the completion of administrative procedures and documents to support non-power modes for the implementation of NFPA 805. The NRC staff concludes that this action is acceptable because the licensee identified a required action that will incorporate the provisions of NFPA 805 in the licensee's fire protection program and included the action as an implementation item in LAR Attachment S, which would be required by the proposed license condition.
3.5.3.4 Conclusion for NPO Based on its review of the information provided in the LAR, the NRC staff concludes that the licensee used methods consistent with the guidance provided in RG 1.205, NEI 04-02, and FAQ 07-0040 to identify the equipment required to achieve and maintain the fuel in a safe and stable condition during NPO modes. Furthermore, the NRC staff concludes that the licensee has a process in place to ensure that fire protection defense-in-depth measures will be implemented to achieve the KSFs during plant outages and that any required actions will be completed through implementation items identified in LAR Attachment S, Table S-2, which would be required by the proposed license condition.
NFPA 805 requires that the nuclear safety performance criteria 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. The licensee has:
Identified the KSFs required to support the nuclear safety performance criteria during non-power operations; Identified the plant operating states where further analysis is necessary during non-power operations;
- 133 -
Identified the SSCs required to meet the KSFs during the plant operating states 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 lost as a direct result of fire-induced damage; and.
Planned/implemented modifications to appropriate station procedures in order to employ one or more fire protection strategies for reducing risk at these pinch points during HREs.
Accordingly, based on the information provided in the LAR as supplemented, and subject to completion of the implementation items as stated in the proposed license condition, the NRC staff concludes that the licensee has provided reasonable assurance that the nuclear safety performance criteria will be met during NPO modes and HREs.
3.5.4 Conclusion for Section 3.5 The NRC staff reviewed the licensee's risk-informed, performance-based fire protection program, as described in the LAR and its supplements, to evaluate the nuclear safety capability assessment results. The licensee used a combination of the deterministic approach and the performance-based approach, in accordance with NFPA 805, Sections 4.2.3 and 4.2.4.
For those fire areas that utilized a deterministic approach, the NRC staff confirmed the following:
The licensee followed appropriate NRC and industry guidance in evaluating compliance to NFPA 805 Chapter 4 on a fire area basis; The engineering evaluations for deviations from the existing fire protection program were evaluated and found to be valid and acceptable 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 nuclear safety performance criteria for each fire area; and The required automatic fire suppression and automatic fire detection systems were appropriately documented for each fire area.
Accordingly, the NRC staff concludes that the licensee has provided adequate documentation that each fire area utilizing the deterministic approach does so in accordance with NFPA 805, Section 4.2.3.
For those fire areas that utilized the performance-based approach, the NRC staff confirmed the following:
- 134 -
The licensee followed appropriate NRC and industry guidance in evaluating compliance to NFPA 805 Chapter 4 on a fire area basis; Fire suppression effects were evaluated and found to have no adverse impact on the ability to achieve and maintain the nuclear safety performance criteria for each fire area; For those fire areas utilizing the fire risk evaluation performance-based approach, all associated variances from deterministic requirements were evaluated using the fire risk evaluation performance-based method to addressrisk impact, defense-in-depth, and safety margin, and found to be acceptable; For those fire areas utilizing the fire modeling performance based approach, all associated variances from deterministic requirements were evaluated using performance based fire modeling and found to be acceptable.
All recovery actions 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 805, Section 4.2.4; and The required automatic fire detection and suppression systems were appropriately documented for each fire area.
Four fire areas were evaluated using the fire modeling performance-based approach in accordance with NFPA 805 Section 4.2.4.1. The review of these fire areas is discussed in SE Section 3.3.
Accordingly, the NRC staff concludes that the licensee has provided adequate documentation that each fire area utilizing the performance-based approach in accordance with NFPA 805, Section 4.2.4 is able to achieve and maintain the nuclear safety performance criteria. The licensee followed appropriate NRC and industry guidance in performing the fire area analyses.
The licensee appropriately addressed recovery actions and evaluated the risk of their use. Taken together, there is reasonable assurance that the associated evaluations meet the requirements for risk, defense-in-depth and safety margins.
The NRC staff's review of the licensee's analysis and outage management process during non-power operational modes concludes that the licensee provided adequate justification that the nuclear safety performance criteria will be met during NPO modes and HREs. The NRC staff review also found that the normal fire protection program defense-in-depth actions are credited for addressing the risk impact of those fires which potentially affect one or more trains of equipment that provide a KSF required during NPO modes, but would not be expected to cause the total loss of that KSF. The NRC staff concludes that this overall approach for fire protection during NPO modes is acceptable because the licensee provided reasonable assurance that the nuclear safety performance criteria will be met during NPO modes and HREs.
3.6 Radioactive Release Performance Criteria
- 135 -
3.6.1 Method of Review NFPA 805 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 in any plant operational mode as follows:
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 Either of the following objectives shall be met during all operational modes and plant configurations.
(1)
Containment integrity is capable of being maintained.
(2)
The source term is capable of being limited.
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.
The NRC staff has endorsed (with certain exceptions) the methodology given NEI 04-02 (Reference 5) as providing methods acceptable to the staff for establishing a fire protection program consistent with NFPA 805 (Reference 1) and 10 CFR 50.48(c) in RG 1.205 (Reference 2). Using these methods, the licensee has assessed the capability of its fire protection program to meet the NFPA 805 performance criteria as contained in NEI 04-02 and FAQ 09-0056 (Reference 60). The results of the licensee's assessment are documented in the LAR (Reference 6).
The NRC reviewed the assessment provided in the LAR in order to determine if the existing fire protection program with its planned modifications would meet the radioactive release performance criteria requirements of a risk-informed, performance-based fire protection program, in accordance with 10 CFR 50.48(a) and (c) using the guidance in RG 1.205 and NUREG-0800, Section 9.5.1.2 (Reference 38). The NRC staff also performed an audit of the licensee's evaluation to determine whether the VCSNS fire protection program and its planned modifications would be capable of meeting the NFPA 805 radioactive release goals, objectives, and performance criteria. The results of the NRC staff evaluation are provided below.
3.6.2 Scope of Review The licensee's evaluation of the capability of the VCSNS fire protection program to meet the goals, objectives, and performance criteria of NFPA 805 was performed for all plant operating modes
- 136 -
(including power and non-power operations) and for all plant areas. The licensee's review as documented in the LAR found that the fire suppression activities, as defined in the pre-fire plans and fire brigade firefighting instruction operating guidelines, were written and valid for any plant operating mode. The NRC staff concludes that the scope of the licensee's assessment was adequate because the review included all modes of plant operation and all plant areas.
3.6.3 Identification of Plant Areas Containing Radioactive Materials The licensee performed a screening of plant fire areas to determine where there was a potential for generating radioactive effluents during firefighting operations. The results of the screening review are documented in LAR Attachment E, Table E-1 "Radioactive Release Transition Engineered Controls Review." The fire areas where there was no possibility of radioactive materials being present were identified and eliminated from further review. Each fire area that had the potential for generation of radioactive effluents created by firefighting activities was identified and "screened in" for further evaluation.
The screened in areas included those areas where most of the radioactive materials were present such as in the Auxiliary Building, Fuel Handling Building, Intermediate Building, Reactor Building, and Radiological Maintenance Building. The review found that these areas had adequate engineered controls for containment of liquid and gaseous effluent. The licensee's review identified the existing engineering controls that were present and sufficient to contain gaseous and liquid effluent. These engineering controls credited for containment of gaseous and liquid effluent are identified and documented in LAR Attachment E. The NRC staff concluded that the identified engineering controls were adequate because they provided sufficient capacity to contain the gaseous and liquid firefighting effluents.
The licensee's review also identified other plant areas where radioactive materials were present and where there were minimal or no engineered controls for containment of effluents. These areas include the Containment Access Building, Control Building, Contaminated Storage Warehouse, Radiography Laboratory, and Warehouse A. For these areas, the potential for radioactive release and radiation exposure to members of the public was evaluated in a qualitative assessment (see SE Sections 3.6.5 and 3.6.6).
The NRG staff concludes that the licensee's screening of plants areas and identification of the potentially affected areas is an adequate assessment because the review incorporated all plant areas, and identified potentially affected areas with and without engineering controls, in accordance with the guidance in NEI 04-02 as endorsed by RG 1.205.
3.6.4 Fire Pre-Plans
. The licensee reviewed its existing fire pre-plans to determine whether its fire protection program is adequate to ensure that gaseous and liquid radioactive effluents generated as a direct result of fire suppression activities would be contained and monitored before release to unrestricted areas.
The results of the licensee's review are documented in the LAR Attachment E, Table E-1 and included the following steps:
Identification of applicable documentation, including fire pre-plans, procedures, and support drawings.
- 137 -
o Review of current documentation to identify whether the current documents discuss the containment and monitoring of potential contamination involving fire suppression activities.
Review of engineering controls for gaseous effluents to determine in which areas the gaseous effluents are contained (for example contaminated smoke and related particulates).
Review of engineering controls for liquid effluents to determine in which areas the liquid effluents are contained (for example automatic or manual fire-fighting water).
An identification of those documents needing revision such as to provide for monitoring and containment of fire suppression agents and radioactive release.
The NRC staff concludes that the licensee's evaluation of the fire pre-plans is adequate because the review was comprehensive and was performed in accordance with the guidance in NEI 04-02, Appendix G, as endorsed by RG 1.205.
3.6.5 Gaseous Effluent Controls In areas where engineering controls exist for containment, filtering, and monitoring of gaseous effluent, the engineering controls were determined to provide adequate containment because the effluent was either contained, or filtered to remove radioactive materials and subsequently monitored prior to discharge.
For plant areas where the effectiveness of the installed engineering controls was not adequate to contain the gaseous effluent, the proposed license condition would require the licensee to modify the fire protection program to establish compensatory actions such that the fire brigade and Radiation Protection personnel will manually establish containment and perform monitoring of radioactive effluent. For these plant areas, the NRC staff concludes that NFPA 805 radioactive release goals, objectives, and performance criteria will be met because the radioactive release will be manually contained to within acceptable limits by a combination of the installed engineered controls and compensatory actions taken by the fire brigade and Radiation Protection personnel.
Plant areas with gaseous effluent release paths that are not considered to be significant are the Control Building, Turbine Building, and Intermediate Building. For these areas, the licensee will use portable instrumentation to monitor these areas in the event of a fire involving radiological contamination. Pre-fire plans and fire brigade training materials will be revised to include precautions and strategies to prevent or minimize cross contamination during fire-fighting activities.
In other plant areas where engineered controls were not sufficient for containment of potentially significant radioactive effluents, the licensee evaluated potential releases using a bounding quantitative analysis. The bounding case was the gaseous effluent from the Hot Warehouse where the largest single radioactive source was identified. The licensee performed a dose
- 138 -
assessment based on the type of radionuclides that are stored, and,the maximum amount of radioactive material that was allowed to be stored, and then assumed to be released during a fire.
During the NRC's audit of the licensee's LAR, the NRC reviewed the licensee's calculations used to perform the bounding analysis:
Calculation TR07800-006, "NFPA 805 Radioactive Release Report."
Calculation DC00040-116, "EAB TEDEs - NFPA 805 Compliance."
The NRC staff reviewed the licensee's assessment and determined the assessment to be adequate because models and assumptions used were conservative and consistent with analytical methods that are recognized by the NRC as acceptable methods.
The NRC staff concludes that the licensee adequately minimized a potential release, and performed a bounding analysis based on the maximum amount of radioactive material that is routinely stored. The NRC staff also concludes that the public dose from radioactive material released as a gaseous effluent during a fire would not exceed of the radiological release performance criteria of NFPA 805 and the public dose limits of 1 O CFR Part 20.
3.6.6 Liquid Effluent Controls The licensee identified those areas where sufficient engineering controls exist for containment of liquid effluent (e.g., floor drains routed to sumps and tanks). The NRC staff reviewed those engineering controls and determined that those controls provided adequate containment because the effluent is collected, stored, processed and monitored in the Radwaste Building prior to discharge.
The licensee's review also identified those areas where there were not sufficient engineered controls to adequately contain potential liquid effluents released during firefighting activities, such as in the Hot Warehouse. In these areas, the licensee identified strategies to mitigate the potential release such as the use of storm drain covers.
For analysis purposes, the licensee performed a bounding evaluation of the worst case liquid release, assumed to be the rupture of the radwaste tank. To mitigate this release, the licensee will revise the fire protection program procedures and training programs to have the fire brigade and Radiation Protection staff trained and instructed to monitor, and if necessary, to install storm drain covers to minimize the release. The assessment concluded that the potential radiological impact of uncontained liquid effluent would not exceed the radiological release performance criteria of NFPA 805 and the public dose limits of 10 CFR Part 20.
The NRC staff reviewed the calculational methods and concludes that the licensee adequately assessed the potential dose impact of uncontained liquid effluent because the bounding assessment was based on conservative assumptions and analytical methods. Based on the NRC staff's review of this bounding assessment, the NRC concludes that a potential radiological liquid effluent release would not exceed the radiological release performance criteria of NFPA 805 and the public dose limits of 10 CFR Part 20.
- 139 -
3.6.7 Fire Brigade Training Materials The licensee reviewed the fire brigade training materials to ensure they were consistent with the pre-fire plans in terms of containment and monitoring of potentially contaminated smoke and fire suppression water. The review is documented in LAR Attachment E.
Each.training module and lesson plan was evaluated, and those training materials needing improvements were identified and documented. The training materials that will be revised are identified in the LAR, Attachment E, and described in LAR Attachment S, Table S-2, Implementation Items 5 and 7, and the NRC staff concludes that these actions are acceptable because the actions will incorporate the provisions of NFPA 805 in the licensee's fire protection program and the actions are included as implementation items in LAR Attachment S, which would be required by the proposed license condition.
The training material revisions will describe the actions the fire brigade will take to ensure that the engineering controls are intact and capable of supporting containment of gaseous and liquid effluents, including the use of manual mitigation methods {e:g., storm drain covers) when necessary.
The NRC staff reviewed the licensee's evaluation of training materials and concludes that upon completion of the implementation items, the training materials will be adequate to instruct the VCSNS staff to implement the fire protection program because plant staff will be informed and capable of taking actions to limit the public dose to within the radiological release performance criteria of NFPA 805.
3.6.8 Conclusion The NRC staff's evaluation is based on:
- 1)
Information and analyses provided in the LAR;
- 2)
Use of installed and manual engineered controls to contain potential releases;
- 3)
Use of fire pre-plans;
- 4)
Use of revised fire brigade response procedures and training procedures; and
- 5)
Bounding dose assessments.
Based on these factors, the NRC staff concludes that the licensee's risk informed, performance based fire protection program 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 likely to exceed the radiological release performance criteria of NFPA 805 and the radiological dose limits in 10 CFR Part 20. The NRC staff concludes that upon completion of the implementation items, the licensee's fire protection program will comply with the
- requirements specified in NFPA 805, Sections 1.3.2, 1.4.2, and 1.5.2 and that this approach is acceptable.
- 140 -
3.7 NFPA 805 Monitoring Program For this SE section, the following requirements from NFPA 805 (Reference 1 ), Section 2.6, are applicable to the NRC staff's review of the LAR:
NFPA 805 Section 2.6: "Monitoring," states that:
A monitoring program shall be established to ensure that the availability and reliability of the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria.
Monitoring shall ensure that the assumptions in the engineering analysis remain valid.
NFPA 805 Section 2.6.1: "Availability, Reliability, and Performance Levels," states that:
Acceptable levels of availability, reliability, and performance shall be established.
NFPA 805 Section 2.6.2: "Monitoring Availability, Reliability, and Performance," states that:
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," states that:
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 (Reference 6) Section 4.6, "Monitoring Program," that the licensee developed to monitor availability, reliability, and performance of fire protection program systems and features after transition to NFPA 805. The focus of the NRC staff review was on critical elements related to the monitoring program, including the selection of fire protection 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 risk-informed, performance-based fire protection program implementation, which the NRC staff concludes is acceptable.,
The licensee stated that it will develop an NFPA 805 monitoring program consistent with FAQ 10-0059 (Reference 61 ). 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, nuclear safety capability
- 141 -
assessment equipment, SSCs relied upon to meet radioactive release criteria, and fire protection programmatic elements.
As described above, NFPA 805, Section 2.6, requires that a monitoring 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 fire protection program 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, is acceptable and assures that the licensee will implement an effective program for monitoring risk significant fires because it:
Establishes the appropriate scope of SSCs 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; and Requires corrective actions when SSC availability, reliability, and performance criteria targets are exceeded in order to 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 safety evaluation, completion of the licensee's NFPA 805 Monitoring Program is an implementation item, addressed in LAR Attachment S, Table S-2, Implementation Item 4.
The NRC staff concludes that completion of the monitoring program on the same schedule as the implementation of NFPA 805 is acceptable because the monitoring program will be completed by December 31, 2015 as described in LAR Attachment S, Table S-2, which is concurrent with completion of the modifications to achieve full compliance with 10 CFR 50.48(c).
- 3. 7.1 Conclusion for Section 3. 7 The NRC staff reviewed the licensee's risk-informed, performance-based fire protection program and concludes that the licensee's approach for meeting the requirements of NFPA 805, Section 2.6, regarding the monitoring program is acceptable and that there is reasonable assurance that the licensee with develop a monitoring program that meets the requirements specified in Section 2.6.1, 2.6.2 and 2.6.3 of NFPA 805 because the licensee identified an action to revise plant documents to monitor and trend the fire protection program, and included that action as an implementation item which would be required by the proposed license condition.
- 142 -
3.8 Program Documentation, Configuration Control, and Quality Assurance For this SE section, the requirements from NFPA 805 (Reference 1 ), Section 2.7, "Program Documentation, Configuration Control and Quality," are applicable to the NRC staff's review of the LAR (Reference 6) in regard to the appropriate content, configuration control, and quality of the documentation used to support the VCSNS fire protection program transition to NFPA 805.
NFPA 805, Section 2.7.1.1, "General," states that:
The analyses performed to demonstrate compliance with this standard shall be documented for each nuclear power plant (NPP). The intent of the documentation is that the assumptions be clearly defined and that the results be easily understood, that results be clearly and consistently described, and that sufficient detail be provided to allow future review of the entire analyses. Documentation shall be maintained for the life of the plant and be organized carefully so that it can be checked for adequacy and accuracy either by an independent reviewer or by the AHJ.
NFPA 805, Section 2.7.1.2, "Fire Protection Program Design Basis Document," states that:
A fire protection program design basis document shall be established based on those documents, analyses, engineering evaluations, calculations, and so forth that define the fire protection design basis for the plant. As a minimum, this document shall include fire hazards identification and nuclear safety capability assessment, on a fire area basis, for all fire areas that could affect the nudear safety or radioactive release performance criteria defined in Chapter 1.
NFPA 805, Section 2.7.1.3, "Supporting Documentation," states that:
Detailed information used to develop and support the principal document shall be referenced as separate documents if not included in the principal document.
NFPA 805, Section 2.7.2.1, "Design Basis Document," states that:
The design basis document shall be maintained up-to-date as a controlled document. Changes affecting the design, operation, or maintenance of the plant shall be reviewed to determine if these changes impact the fire protection program documentation.
NFPA 805, Section 2.7.2.2, "Supporting Documentation," states that:
Detailed supporting information shall be retrievable records. Records shall be revised as needed to maintain the principal documentation up-to-date.
- 143 -
NFPA 805, Section 2. 7.3.1, "Review," states that:
Each analysis, calculation, or evaluation performed shall be independently reviewed.
NFPA 805, Section 2.7.3.2*, "Verification and Validations" states that:
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," states that:
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," states that:
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" states that:
An uncertainty analysis shall be performed to provide reasonable* assurance that the performance criteria have been met.
3.8.1 Documentation The NRC staff reviewed LAR Section 4.7.1, "NFPA 805 Documentation Requirements (NFPA 805, Section 2.7.1)," to evaluate the VCSNS fire protection program design basis document and supporting documentation.
The VCSNS design basis is a compilation of multiple documents (i.e., fire safety analyses, calculations, engineering evaluations, nuclear safety capability assessments, etc.), databases, and drawings which are identified in LAR Figure 4-8, "NFPA 805 Transition - Planned Post-Transition Documentation Relationships." The licensee stated that the analyses conducted to support the NFPA 805 transition were performed in accordance with VCSNS processes which meet or exceed the requirements for documentation outlined in NFPA 805, Section 2.7.1.
Specifically, the licensee stated that 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 licensee further stated that the process includes provisions for appropriate design and engineering review and approval and that the approved analyses are considered controlled documents and are accessible via the
- 144 -
licensee's document control system. The licensee further stated that the analyses 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 fire protection program will be maintained for the life of the plant (as required by NFPA 805) 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 fire protection program 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 fire protection program design basis documentation, is acceptable.
3.8.2 Configuration Control The NRC staff reviewed LAR Section 4.7.2, "NFPA 805 Configuration Management(NFPA 805, Sections 2.2.9/2.7.2)," in order to evaluate the VCSNS configuration control process for the new NFPA 805 fire protection program.
To support the many other technical, engineering and licensing programs, the licensee has existing configuration control processes and procedures for establishing, revising, or utilizing program documentation. Accordingly, the licensee is integrating the new fire protection program 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 VCSNS licensing programs, including the fire protection program.
The LAR stated that the configuration control process includes provisions for appropriate design, engineering reviews and approvals, and that approved analyses are considered controlled documents available through the licensee's document control system. The LAR also stated that analyses based on the PRA program, which includes the fire risk evaluations, 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 36).
Configuration control of the existing fire protection program during the transition period is maintained by the change evaluation process, as defined in existing configuration management and configuration control procedures. LARAttachment S, Table S-2 includes Implementation Items 13 and 14 to update engineering and fire PRA procedures to manage configuration control of NFPA 805 analysis documents. The NRC staff concludes that these actions are acceptable because they are included as implementation items in LAR Attachment S, Table S-2, which would.
be required by the proposed license condition.
The NRC staff reviewed the licensee's process for updating and maintaining the Fire PRA in order to reflect plant changes made after completion of the transition to NFPA 805 in SE Section 3.4.
- 145 -
Based on the description of the VCSNS configuration control process, which indicates that the new fire protection program design basis and supporting documentation will be controlled documents and that plant changes will be reviewed for impact on the fire protection program, which would be required by the proposed license condition, the NRC staff concludes that there is adequate justification that the requirements of NFPA 805, Sections 2.7.2.1 and 2.7.2.2 will be met.
3.8.3 Quality The NRC staff reviewed LAR Section 4.7.3, "NFPA 805 Quality Requirements (NFPA 805, Section 2.7.3)," to evaluate the quality of the engineering analyses used to support transition of the VCSNS fire protection program 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 fire protection program, as appropriate.
3.8.3.1 Review NFPA 805, Section 2.7.3.1 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 existing 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 805, Section 2.7.3.1, is acceptable.
3.8.3.2 Verification and Validation (V&V)
NFPA 805, Section 2.7.3.2 requires that each calculational model or numerical method used be verified and validated (V&V) through comparison to test results or other acceptable models. The 1.icensee stated that the calculational models and numerical methods used in support of the transition to NFPA 805 were V&V, and that the calculational models and numerical methods used post-transition will be similarly V&V. As an example, the licensee provided extensive information related to the V&V of fire models used to support the development of the VCSNS fire risk evaluations and analyses used to support the fire modeling performance-based approach. 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 48), documents the V&V of five selected fire models commonly used to support applications of risk-informed, performance-based fire protection at nuclear power plants. The seven volumes of this NU REG-series 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
- 146 -
considering the identified uncertainties, these models may be employed to demonstrate compliance with the requirements of 10 CFR 50.48(c) as part of an approved performance-based approach in accordance with NFPA 805, Chapter 4.
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, the NRC staff concludes that the use of these models is acceptable, provided that the intended application is within the appropriate limitations, as identified in NUREG-1824.
Table 3.8-1, "V&V Basis for Fire Modeling Correlations Used at VCSNS," in SE Attachment A and Table 3.8-2, "V&V Basis for Other Fire Models and Related Calculations Used at VCSNS in SE Attachment B, identify these empirical correlations and fire models, respectively, as well as a staff disposition for each.
The fire modeling employed by the licensee in the development of the fire risk evaluations and fire modeling performance-based approach used empirical correlations that provide bounding solutions for the zone of influence; and conservative input parameters, which produced conservative results for the fire modeling analysis. See SE Sections 3.3 and 3.4.2.3 for further discussion of the licensee's fire modeling method.
3.8.3.2.2 Discussion of RAI Responses In letters dated July 26, 2012 (Reference 20), and August 13, 2013 (Reference 21), the NRC staff requested additional information from the licensee. In letters dated October 10, 2012-(Reference 8), February 1, 2013 (Reference 9), April 1, 2013 (Reference 10), and October 14, 2013 (Reference 11 ), the licensee provided its responses to the requests.
During the audit, the NRC staff noted that the fire dynamic tools used were implemented in a proprietary fire modeling database. In FM RAI 03.a (Reference 20), the NRC staff requested that the licensee explain how the database and built-in FDT calculations were verified. In its response to FM RAI 03.a (Reference 8), the licensee stated that a V&V study was completed for the main quantification module of the database and that the study involved a comparison between the ZOI reported in the database and hand calculations for two test cases.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated that the fire dynamic tool calculations were appropriately verified.
In FM RAI 03.b (Reference 20), the NRC staff requested that the licensee demonstrate that the fire models used in support of the fire modeling approach (NFPA 805 Section 4.2.4.1) were applied within the validation range of input parameters or to justify the application of the model outside the validation range reported in NUREG-1824 or other V&V basis documents. In its response to FM RAI 03.b (Reference 8), the licensee stated that for all compartments in which fire modeling was used, the dimensionless parameters that describe the fire scenarios were calculated and compared against the validation range specified in
- 147 -
NUREG-1824 and that when parameters fell outside the validation range, a sensitivity case was run to show that the use of CFAST in the current application was justified.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated that the fire models used in support of the fire modeling approach were either applied within their validation range or that their use outside their validation range was justified.
In FM RAI 03.c (Reference 20), the NRC staff requested that the licensee confirm that the normalized flame height values in LAR Attachment J account for the elevation of the fire, and if. not, to recalculate the normalized parameter and confirm that the new value is within the parameter's range of applicability or provide justification for the use of the fire model in cases when the new parameter value is outside the validation range. In its response to FM RAI 03.c (Reference 8), the licensee stated that it had determined that the normalized flame height values did not account for the fire elevation and that recalculating the normalized parameters showed that the values were within the validation range for CB10, CB12, and CB18. The licensee further stated that for 1811, a sensitivity study was conducted to show that bringing the normalized parameter inside the range does not change the conclusions of the CF AST analysis.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated that flame height values were within the validated range for three of the fire areas, and that for the remaining fire area, the use of a flame height value outside of the validated range was justified.
In FM RAI 03.d (Reference 20), the NRC staff requested that the licensee demonstrate that the fire models used in support of the fire risk evaluations were applied within their validation range, or to provide technical justification for their use outside the validation range. In its response to FM RAI 03.d (Reference 8), the licensee explained that for all compartments in which fire modeling was used, the normalized parameters that describe the fire scenarios were calculated and compared against model validation parameters, as specified in NUREG-1824 and that when parameters fell outside the validation range, a sensitivity case was run with CFAST to demonstrate that the conclusions were unchanged and to show that the use of CFAST in the current application is justified.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated that the fire models used in support of the fire risk evaluations were applied within their validation range, or provided appropriate technical justification for their use outside the validation range.
In FM RAI 03.e (Reference 20), the NRC staff requested that the licensee provide the origin of the version of Heskestad's plume temperature equation that was used in the hand calculations reviewed during the audit, and to explain how this equation was verified and validated. In its response to FM RAI 03.e (Reference 8),
the licensee explained that the source of the plume temperature correlation used
3.8.3.2.3
- 148 -
in the calculations reviewed during the audit is listed in the spreadsheet as Heskestad's chapter on fire plumes, flame height, and air entrainment in the Society of Fire Protection Engineering (SFPE) Handbook of Fire Protection Engineering (Reference 94). The licensee further stated that the plume temperature correlation is validated in NUREG-1824 (Reference 48), as a part of the FIVE Rev 1 software package.
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated that it properly verified and validated the plume temperature correlation used in the calculations.
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 fire protection program changes, including those for V&V. Revision of the applicable post-transition processes and procedures to include NFPA 805 requirements for V&V are included in LAR Attachment S, _
Table S-2 as Implementation Items 1, 2,4, 7, 13, 14, 15, 17, and 18 and the NRC staff concludes that these actions are acceptable because they will incorporate the NFPA 805 requirements for V&V into the licensee's processes and procedures and because they are included as implementation items which would be required by the proposed license condition.
3.8.3.2.4 Conclusion for Section 3.8.3.2 Based on the licensee's description of the VCSNS 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 because the models are consistent with approved uses in NRC guidance or other authoritative publications and the licensee has identified actions that will result in compliance with NFPA 805 and those actions would be required by the proposed license condition.
The NRC staff concludes that the licensee's approach provides adequate justification that the fire modeling used in the development of the fire scenarios for the fire risk evaluations and fire modeling performance-based approach is appropriate, and thus acceptable for use in transition to NFPA 805 because the V&V of the empirical correlations used by the licensee are consistent with either NUREG-1824 or the SFPE Handbook of Fire Protection Engineering.
3.8.3.3 Limitations of Use NFPA 805, Section 2.7.3.3 requires that only acceptable engineering methods and numerical models be used for transition to the extent that these 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.
- 149 -
3.8.3.3.1 General The NRC staff assessed the acceptability of empirical correlations and fire models.in terms of the limits of its use. SE Table 3.8-1 in SE Attachment A and SE Table 3.8-2, in SE Attachment B, summarize the correlations and fire models used, how each was applied in the VCSNS fire risk evaluations, the V&V basis for each, and the NRC staff evaluation for each.
3.8.3.3.2 Discussion of RAls In letters dated July 26, 2012 (Reference 20), and August 13, 2013 (Reference 21 ), the NRC staff requested additional information from the licensee. In letters dated October 10, 2012 (Reference 8), February 1, 2013 (Reference 9), April 1, 2013 (Reference 10), and October 14, 2013 (Reference 11 ), the licensee provided its responses to the requests.
In FM RAI 04 (Reference 20), the NRC staff requested that the licensee identify uses, if any, of the fire modeling tools outside the limits of applicability of the method, and to explain for those cases how the use of the fire model was justified.
In its response to FM RAI 04 (Reference 8), the licensee explained that the use of fire modeling tools was justified by the method described in NUREG-1934 (Reference 81 ). The licensee further stated that dimensionless parameters were compared against the V&V applicability ranges for the various models and that when the dimensionless parameters were outside of the range, a sensitivity calculation was performed to demonstrate the appropriateness of using the model for the application.
3.8.3.3.3 The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated justification for the use of fire modeling tools in accordance with N*Rc endorsed guidance.
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 fire protection program 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 are described in LAR Attachment S, Table S-2, Implementation Items 1, 2, 4, 7, 13, 14, 15, 17, and 18 and the NRC staff considers these actions acceptable because they will incorporate the NFPA 805 requirements for limitations of use into the licensee's processes and procedures and because they are included as implementation items which would be required by the proposed license condition.
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 fire risk evaluations were generally used within their limitations, and the description of the VCSNS 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 because the models are consistent with approved uses in NRC guidance or
- 150 -
other authoritative publications and the licensee has identified actions that will result in compliance with NFPA 805 and those actions would be required by the proposed license condition.
3.8.3.4 Qualification of Users NFPA 805, Section 2.7.3.4 requires that personnel performing engineerin*g analyses and applying numerical methods (e.g., fire modeling) 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 personr:iiel 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, the requirements in NFPA 805, Section 2.7.3.4 are being addressed through the implementation of an engineering qualification process at VCSNS. The licensee has developed qualification and training requirements for personnel performing engineering analyses and numerical methods. Appropriate qualification for licensee's staff and consultants to use and apply fire modeling tools include required reading of technical guidance material, classroom training in fire analysis and fire modeling, demonstration of comprehension and proficiency forfire modeling.
This qualification and training process is addressed in a training module worksheet in accordance with a site engineering personnel training program.
The NRC staff concludes that appropriately competent and experienced personnel developed the fire risk evaluations and fire modeling performance-based approach, including the supporting fire modeling calculations and including the additional documentation for models and empirical correlations not identified in previous NRG-approved V&V documents.
The post-transition qualification trai,ning program will be implemented to include NFPA 805 requirements for Qualification of Users and is included in LAR Attachment S, Table S-2 as part of Implementation Item 17. The NRC staff concludes that this action is acceptable because the action will incorporate the NFPA 805 requirements for qualifications of users into the licensee's training program and because the action is part of an implementation item that would be required by the proposed license condition.
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, Section 2.7.3.5 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
- 151 -
analyses.
3.8.3.5:1 General The industry consensus standard for PRA development, i.e., the ASME/ANS PRA standard (Reference 36) includes requirements to address uncertainty. Accordingly, the licensee addressed uncertainty as a part of the development of the fire risk evaluations. 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 50) there are three types of uncertainty associated with fire modeling calculations:
(1)
(2)
(3) 3.8.3.5.2 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 ttie results of the fire modeling analysis.
Model Uncertainty: Idealizations of physical phenomena lead to simplifying assumptions in the formulation of the model equations. In addition, the numerical solution of 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 81)
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.
Discussion of RAls In letters dated July 26, 2012 (Reference 20), and August 13, 2013 (Reference 21), the NRC staff requested additional information from the licensee. In letters dated October 10, 2012 (Reference 8), February 1, 2013 (Reference 9), April 1, 2013 (Reference 10), and October 14, 2013 (Reference 11 ), the licensee provided its responses to the requests.
In FM RAI 06.a (Reference 20), the NRC staff requested that the licensee describe how the uncertainty associated with the fire model input parameters was accounted for. In its response to FM RAI 06.a (Reference 8), the licensee explained that parameter uncertainty is accounted for by using bounding heat release rates, ignoring fuel burnout in cable tray fires postulated as secondary combustibles, and selection of the scenario with the largest number of panels and cable trays for the CFAST calculations.
- 152 -
The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee demonstrated that uncertainty associated with the fire model input parameters is appropriately accounted for.
In FM RAI 06.b (Reference 20), the NRC staff requested that the licensee describe how the "model" and "completeness" uncertainty were accounted for. In its response to FM RAI 06.b (Reference 8), the licensee explained that the detailed fire modeling reports for individual fire areas include a discussion of margin based on the model uncertainties (biases) listed in NUREG-1824 and that completeness uncertainties were handled with sensitivity studies. The licensee further stated that sensitivity calculations were performed for those fire scenarios with normalized parameters outside of the validation range for CFAST.
The NRC staff concludes that the licensee's response to the RAls is acceptable because the licensee demonstrated that model and completeness uncertainty are appropriately accounted for.
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 fire protection program changes, including those regarding uncertainty analysis. Revision of the applicable post-transition processes and procedures to include NFPA 805 requirements regarding uncertainty analysis are included in LAR Attachment S, Table S-2 as Implementation Items 1, 2, 4, 7, 13, 14, 15, 17, and 18 and the NRC staff considers these actions acceptable because they will incorporate the NFPA 805 quality requirements regarding uncertainty anlaysis into the licensee's processes and procedures and because they are included as implementation items which would be required by the proposed license condition.
3.8.3.5.4 Conclusion for Section 3.8.3.5 Based on the licensee's description of the 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 and subject to completion of the implementation items as stated in the proposed license condition, the NRC staff concludes that the risk-informed, performance-based fire protection quality assurance process meets each of the requirements of NFPA 805, Section 2.7.3, which includes 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.
3.8.4 Fire Protection Quality Assurance Program GDC 1 of Appendix A to 10 CFR Part 50 requires the following:
- 153 -
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 quality assurance program in accordance with the guidelines of NUREG-0800, Section 9.5.1 position C.4, "Quality Assurance Program,"
(Reference 78). In addition, the guidance in NEI 04-02 (Reference 5) Appendix C suggests that the LAR include a description of how the existing fire protection quality assurance program will be transitioned to the new NFPA 805 risk-informed, performance-based fire protection program, as discussed below.
The LAR states that the fire protection quality assurance program will be integrated into and implemented by the VCSNS nuclear quality assurance program, although certain aspects of that program are not applicable to the fire protection program. Further, the licensee will revise the quality assurance program to reflect the applicable requirements ofNFPA 805 Section 2.7.3 and has included those revisions in LAR Attachment S, Table S-2, as Implementation Items 1, 2, 4, 7, 13, 14, 15, 17, and 18 and the NRC staff considers these actions acceptable because they are included as implementation items which would be required by the proposed license condition.
The NRC staff concludes that the licensee's changes to the fire protection quality assurance 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 quality assurance program that are required by NFPA 805 transition and post-transition, and have identified required actions to do so that are included in the proposed license condition.
3.8.5 Conclusion for Section 3.8 The NRC staff reviewed the licensee's risk-informed, performance-based fire protection program as described in the LAR, as supplemented, to evaluate the NFPA 805 program documentation content, the associated configuration control processes, and the appropriate quality assurance requirements. The NRC staff concludes that, subject to completion of the implementation items related to the quality assurance program, as stated in the proposed license condition, the licensee's approach for meeting the requirements specified in NFPA 805 Section 2. 7 is acceptable.
4.0 FIRE PROTECTION LICENSE CONDITION The licensee proposed a fire protection program license condition regarding transition to an risk-informed, performance-based fire protection program under NFPA 805, in accordance with 1 O 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 fire protection program license condition is consistent with the standard fire protection license condition, incorporates all of the relevant features of the transition to NFPA 805 at Virgil C. Summer Nuclear Station, Unit 1, and is, therefore, acceptable.
- 154 -
The following license condition is included in the revised license and will replace Operating License No. NPF-12, Condition 2.C.(18):
Fire Protection Program South Carolina Electric and Gas Company 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 11/15/11, (and supplements dated 1/26/12, 10/10/12, 2/1/13, 4/1/13, 10/14/13, 11/26/13, 1/9/14, 2/25/14, 5/2/14, 5/11/14, 8/14/14, 10/9/14, and 12/11/14) and as approved in the safety evaluation dated 02/11 /15. 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.
- a.
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.
- 1.
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.
- 2.
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 1x10-8/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.
- b.
Other Changes that May Be Made Without Prior NRC Approval
- 1.
Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program
- 155 -
Prior NRC review and approval are not required for chang~s 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 3.1 O); 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.
J
- 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 dated February 11, 2015, 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.
(
- 156 -
- c.
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, "P.lant Modifications Committed," of SCE&G letter RC-14-0196, dated December 11, 2014, by the end of the calendar year 2015. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.
- 3.
The licensee.shall implement items listed in Attachment S, Table S-2, "Implementation Items," of SCE&G letter RC-14-0196, dated December 11, 2014, to complete the transition to full compliance with 10 CFR 50.48(c) by March 31, 2016 as follows:
- a.
Items 3, 6, 7, 8, 10, 13, 14, 17, 19, and 21within180 days of NRC approval.
- b.
Items 1, 2, 4, 11, and 12 by December 31, 2015.
- c.
Items 5, 15, 16, 18, 20, 22, and 23 by March 31, 2016.
5.0
SUMMARY
The NRC staff reviewed the licensee's application, as supplemented by various letters, to transition to a risk-informed, performance-based fire protection program 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 an risk-informed, performance-based fire protection program in accordance with 10 CFR 50.48(c).
Implementation of the risk-informed, performance-based fire protection program in accordance with 10 CFR 50.48(c) will include the application of a new fire protection license condition. The new license condition includes a list of implementation items that must be completed in order to support the conclusions made in this SE, as well as an established date by which full compliance with 10 CFR 50.48(c) will be achieved. Before the licensee is able to fully implement the transition to a fire protection program based on NFPA 805 and apply the new fire protection license condition, to its full extent, the implementation items must be completed within the timeframe specified.
6.0 STATE CONSULTATION
In accordance with the Commission's regulations, the South Carolina State official was notified of the proposed issuance of the amendments on January 21, 2015. The State official had no comments.
- 157 -
7.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves 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 amendment involves no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on August 14, 2012 (77 FR 48561 ). 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 amendments will not be inimical to the common defense and security or to the health and safety of the public.
9.0 REFERENCES
- 1.
National Fire Protection Association, NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition, Quincy, Massachusetts.
- 2.
U. S. Nuclear Regulatory Commission, Regulatory Guide 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants,"
Revision 1, December 2009 (ADAMS Accession No. ML092730314).
- 3.
U.S. Nuclear Regulatory Commission, SECY-98-058, "Development of a Risk-Informed, Performance-Based Regulation for Fire Protection at Nuclear Power Plants," March 1998 (ADAMS Accession No. ML992910106).
- 4.
U.S. Nuclear Regulatory Commission, SECY-00-0009, "Rulemaking Plan, Reactor Fire Protection Risk-Informed, Performance-Based Rulemaking," January 2000 (ADAMS Accession No. ML003671923).
- 5.
Nuclear Energy Institute, NEI 04-02, Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c), Revision 2, Washington, DC April 2008 (ADAMS Accession No. ML081130188).
- 158 -
- 6.
Gatlin, Thomas, South Carolina Electric and Gas Company, letter to U.S. Nuclear Regulatory Commission, "Virgil C. Summer Nuclear Station Unit 1, Docket No. 50-395, Operating License No. NPF-12, License Amendment Request to Adopt NFPA 805 Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition)," dated November 15, 2011 (ADAMS Accession No. ML113320227).
- 7.
Gatlin, Thomas, South Carolina Electric and Gas Company, letter to U.S. Nuclear Regulatory Commission, "Virgil C. Summer Nuclear Station Unit 1, Docket No. 50-395, Operating License No. NPF-12, License Amendment Request to Adopt NFPA 805 Response to Request for Supplemental Information," dated January 26, 2012 (ADAMS Accession No. ML12031A149).
- 8.
Gatlin, Thomas, South Carolina Electric and Gas Company, letter to U.S. Nuclear Regulatory Commission, "Virgil C. Summer Nuclear Station Unit 1, Docket No. 50-395, Operating License No. NPF-12, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information," dated October 10, 2012 (ADAMS Accession No. ML12297A218).
- 9.
Gatlin, Thomas, South Carolina Electric and Gas Company, letter to U.S. Nuclear Regulatory Commission, "Virgil C. Summer Nuclear Station Unit 1, Docket No. 50-395, Operating License No. NPF-12, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information," dated February i, 2013 (ADAMS Accession No. ML13037A111 ).
- 10.
Gatlin, Thomas, South Carolina Electric and Gas Company, letter to U.S. Nuclear Regulatory Commission, "Virgil C. Summer Nuclear Station Unit 1, Docket No. 50-395, Operating License No. NPF-12, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information," dated April 1, 2013 (ADAMS Accession No. ML13092A333).
- 11.
Gatlin, Thomas, South Carolina Electric and Gas Company, letter to U.S. Nuclear Regulatory Commission, "Virgil C. Summer Nuclear Station Unit 1, Docket No. 50-395, Operating License No. NPF-12, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information," dated October 14, 2013 (ADAMS Accession No. ML13289A194).
- 12.
Gatlin, Thomas, South Carolina Electric and Gas Company, letter to U.S. Nuclear Regulatory Commission, "Virgil C. Summer Nuclear Station Unit 1, Docket No. 50-395, Operating License No. NPF-12, License Amendment Request to Adopt NFPA 805
- Response to Request for Additional Information," dated November 26, 2013 (ADAMS Accession No. ML13333A282).
- 13.
Gatlin, Thomas, South Carolina Electric and Gas Company, letter to U.S. Nuclear Regulatory Commission, "Virgil C. Summer Nuclear Station Unit 1, Docket No. 50-395, Operating License No. NPF-12, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information," dated January 9, 2014 (ADAMS Accession No. ML14013A074).
- 159 -
- 14.
Gatlin, Thomas, South Carolina Electric and Gas Company, letter to U.S. Nuclear Regulatory Commission, "Virgil C. Summer Nuclear Station Unit 1, Docket No. 50-395, Operating License No. NPF-12, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information," dated February 25, 2014 (ADAMS Accession No. ML14063A455).
- 15.
Gatlin, Thomas, South Carolina Electric and Gas Company, letter to U.S. Nuclear Regulatory Commission, "Virgil C. Summer Nuclear Station Unit 1, Docket No. 50-395, Operating License No. NPF-12, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information," dated May 2, 2014 (ADAMS Accession No. ML14125A274).
- 16.
Gatlin, Thomas, South Carolina Electric and Gas Company, letter to U.S. Nuclear Regulatory Commission, "Virgil C. Summer Nuclear Station Unit 1, Docket No. 50-395, Operating License No. NPF-12, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information," dated May 11, 2014 (ADAMS Accession No. ML14133A410).
- 17.
Gatlin, Thomas, South Carolina Electric and Gas Company, letter to U.S. Nuclear Regulatory Commission, "Virgil C. Summer Nuclear Station Unit 1, Docket No. 50-395, Operating License No. NPF-12, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information," dated August 14, 2014 (ADAMS Accession No. ML14227A737).
- 18.
Gatlin, Thomas, South Carolina Electric and Gas Company, letter to U.S. Nuclear Regulatory Commission, "Virgil C. Summer Nuclear Station Unit 1, Docket No. 50-395, Operating License No. NPF-12, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information," dated October 9, 2014 (ADAMS Accession No. ML14287A230).
- 19.
Gatlin, Thomas, South Carolina Electric and Gas Company, letter to U.S. Nuclear Regulatory Commission, "Virgil C. Summer Nuclear Station Unit 1, Docket No. 50-395, Operating License No. NPF-12, License Amendment Request to Adopt NFPA 805 Response to Request for Additional Information, dated December 11, 2014 (ADAMS Accession No. ML14350A217).
- 20.
Letter from NRC (Robert E. Martin) to SCE&G (Thomas D. Gatlin), Virgil C. Summer Nuclear Station, Unit No. 1 - Request For Additional Information (TAC NO. ME7586),
July 26, 2012 (ADAMS Accession No. ML12202A027).
- 21.
Letter from NRC (Robert E. Martin) to SCE&G (Thomas D. Gatlin), Virgil C. Summer Nuclear Station, Unit No. 1 - Request For Additional Information (TAC NO. ME7586),
dated August 13, 2013 (ADAMS Accession No. ML13218A195).
- 22.
Letter from NRC (Robert E. Martin) to SCE&G (Thomas D. Gatlin), Virgil C. Summer Nuclear Station, Unit No. 1 - Request For Additional Information (TAC NO. ME7586),
dated August 28, 2013 (ADAMS Accession No. ML13234A442).
- 160 -
- 23.
Letter from NRC (Shawn Williams) to SCE&G (Thomas D. Gatlin), Virgil C. Summer Nuclear Station, Unit No. 1 - Request For Additional Information (TAC NO. ME7586),
dated November 7, 2013 (ADAMS Accession No. ML133088800).
- 24.
Letter from NRC (Shawn Williams) to SCE&G (Thomas D. Gatlin), Virgil C. Summer Nuclear Station, Unit No. 1 - Request For Additional Information (TAC NO. ME7586),
dated December 19, 2013 (ADAMS Accession No. ML133478049).
- 25.
Letter from NRC (Shawn Williams) to SCE&G (Thomas D. Gatlin), Virgil C. Summer Nuclear Station, Unit No. 1 - Request For Additional Information (TAC NO. ME7586),
dated March 24, 2014 (ADAMS Accession No. ML14071A169).
- 26.
Letter from NRC (Shawn Williams) to SCE&G (Thomas D. Gatlin), Virgil C. Summer Nuclear Station, Unit No. 1 - Request For Additional Information Regarding License Amendment Request to Adopt National Fire Protection Association Standard 805 (TAC NO. ME7586), dated July 11, 2014 (ADAMS Accession No. ML14182A473).
- 27.
Schwencer, A., U.S. Nuclear Regulatory Commission, letter to Nichols, T.C., South Carolina Electric and Gas Company, Safety Evaluation Report, Re: Operating License Application and Notice of Federal Register Publication, February 18, 1981 (ADAMS Legacy Library Accession No. 8102260804).
- 28.
Youngblood, 8.J., U.S. Nuclear Regulatory Commission, letter to Nichols, T.C., South Carolina Electric and Gas Company, Issuance of Supplement No. 3 to the Virgil C.
Summer Nuclear Station Safety Evaluation Report, dated January 8, 1982 (ADAMS Legacy Library Accession No. 8202110225).
- 29.
Youngblood, 8.J., U.S. Nuclear Regulatory Commission, letter to Dixon, O.W., South Carolina Electric and Gas Company, Issuance of Amendment No. 1 to Facility Operating License NPF-12 Virgil C. Summer Nuclear Station, Unit No: 1, dated August 20, 1982 (ADAMS Legacy Library Accession No. 8209230570).
- 30.
Hopkins, Jon, 8., U,S. Nuclear Regulatory Commission, letter to Nauman, D.A., South Carolina Electric and Gas Company, V.C. Summer Nuclear Station -Appendix R Reanalysis, dated May 22, 1986 (ADAMS Legacy Library Accession No. 8606110085).
- 31.
Hopkins, Jon, 8., U.S. Nuclear Regulatory Commission, letter to Nauman, D.A., South Carolina Electric and Gas Company, V.C. Summer Nuclear Station - Appendix R Reanalysis, Safety Evaluation of Deviation From Section 111.L.2.D of Appendix R, dated.
November 26, 1986, (ADAMS Accession No. ML091340247).
- 32.
Hopkins, Jon, 8., U.S. Nuclear Regulatory Commission, letter to Nauman, D.A., South Carolina Electric and Gas Company, V.C. Summer Nuclear Station, Appendix R Reanalysis, (TAC No. 57853), Safety Evaluation by the Office of Nuclear Reactor Regulation, Evaluations of Deviations from the Fire Protection Guidelines, dated July 27, 1987, (ADAMS Accession No. ML15030A099)
- 161 -
- 33.
Nuclear Energy Institute, NEI 00-01, "Guidance for Post Fire Safe Shutdown Circuit Analysis," Revision 2, Nuclear Energy Institute (NEI), Washington, DC, May 2009 (ADAMS Accession No. ML091770265).
- 34.
U.S. Nuclear Regulatory Commission, Regulatory Guide 1.17 4, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, May 2011 (ADAMS Accession No. ML100910006).
- 35.
U.S. Nuclear Regulatory Commission, Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, March 2009 (ADAMS Accession No. ML090410014).
- 36.
American Society of Mechanical Engineers (ASME) and American Nuclear Society (ANS) standard 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," dated February 2, 2009.
- 37.
U.S. Nuclear Regulatory Commission, Regulatory Guide 1.189, "Fire Protection for Nuclear Power Plants," Revision 2, October 2009 (ADAMS Accession No. ML092580550).
- 38.
U.S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Chapter 9.5.1.2, "Risk-Informed, Performance-Based Fire Protection Program," Revision 0, December 2009 (ADAMS Accession No. ML092590527).
- 39.
U.S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Chapter 19.1, "Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed License Amendment Requests After Initial Fuel Load," Revision 3, September 2012 (ADAMS Accession No. ML12193A107).
- 40.
U.S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Chapter 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance," Revision 0, June 2007 (ADAMS Accession No. ML071700658).
- 41.
U.S. Nuclear Regulatory Commission, NUREG/CR-6850, "EPRl/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, Volume 1: Summary and Overview,"
September 2005 (ADAMS Accession No. ML052580075).
- 42.
U.S. Nuclear Regulatory Commission, NUREG/CR-6850, "EPRl/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, Volume 2: Detailed Methodology," September 2005 (ADAMS Accession No. ML052580118).
- 162 -
- 43.
U.S. Nuclear Regulatory Commission, NUREG/CR-6850, Supplement 1, "Fire Probabilistic Risk Assessment Methods Enhancements," September 2010 (ADAMS Accession No. ML103090242).
- 44.
Correia, R. P., memorandum to Joseph G. Giitter, U.S. Nuclear Regulatory Commission, "Interim Technical Guidance on Fire-Induced Circuit Failure Mode Likelihood Analysis,"
dated June 14, 2013 (ADAMS Accession No. ML13165A194).
- 45.
U.S. Nuclear Regulatory*Commission NUREGlCR-6931, "Cable Response to Live Fire (CAROL-FIRE)," Volumes 1, 2, and 3, April 2008 (ADAMS Accession Nos. ML081190230, ML081190248, and ML081190261).
- 46.
U.S. Nuclear Regulatory Commission, NUREG/CR-7100, "Direct Current Electrical Shorting in Response to Exposure Fire (DESIREE-Fire): Test Results," April2012 (ADAMS Accession No. ML121600316).
- 47.
U.S. Nuclear Regulatory Commission, NUREG-1805, "Fire Dynamics Tools (FDTS):
Quantitative Fire Hazard Analysis Methods for the U.S. Nuclear Regulatory Commission Fire Protection Inspection Program," December 2004 (ADAMS Accession No. ML043290075).
- 48.
U.S. Nuclear Regulatory Commission, NUREG-1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications," May 2007. Volume 1: Main Report, Volume 2: Experimental Uncertainty, Volume 3: Fire Dynamics Tools (FDTs),
Volume 4: Fire-Induced Vulnerability Evaluation (FIVE-Rev1 ), Volume 5: Consolidated Fire Growth and Smoke Transport Model (CFAST), Volume 6: MAGIC, and Volume 7:
Fire Dynamics Simulator (ADAMS Accession Nos. ML071650546, ML071730305, ML071730493, ML071730499, ML071730527, ML071730504, ML071730543, respectively).
- 49.
U.S. Nuclear Regulatory Commission, NUREG/CR-7010, Volume 1, "Cable Heat Release, Ignition, and Spread in Tray Installations during Fire (CHRISTI FIRE), Phase 1: Horizontal Trays," July 2012 (ADAMS Accession No. ML12213A056).
- 50.
U.S. Nuclear Regulatory Commission, NUREG-1855, Volume 1, NUREG-1855, Vol. 1, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," March 2009 (ADAMS Accession No. ML090970525).
- 51.
U.S. Nuclear Regulatory Commission, NUREG-1921, "EPRl/NRC-RES Fire Human Reliability Analysis Guidelines," July 2012 (ADAMS Accession No. ML12216A104).
51A.
U.S. Nuclear Regulatory Commission, Generic Letter 2006-03, "Potentially Nonconforming Hemyc and MT Fire Barrier Configurations," dated April 10, 2006 (ADAMS Accession No. ML053620142).
/
- 163 -
- 52.
Klein, Alexander R., U.S. Nuclear R~gulatory Commission, memorandum to file, "Close-out of National Fire Protection Association Frequently Asked Question 07-0030 on Establishing Recovery Actions," dated February 4, 2011 (ADAMS Accession No. ML110070485).
- 53.
Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of National Fire Protection Association Frequently Asked Question 07-0038 on Lessons Learned on Multiple Spurious Operations," dated February 3, 2011 (ADAMS Accession No. ML110140242).
- 54.
Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of National Fire Protection Association Standard 805 Frequently Asked Question 07-0039 Incorporation of Pilot Plant Lessons Learned-Table B-2," dated January 15, 2010 (ADAMS Accession No. ML091320068).
- 55.
Nuclear Energy Institute, NEI 00-01, "Guidance for Post Fire Safe Shutdown Circuit Analysis, Revision 1, Nuclear Energy Institute (NEI), Washington, DC, January 2005 (ADAMS Accession No. ML050310295).
- 56.
Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of National Fire Protection Association 805 Frequently Asked Question 07-0040 on Non-Power Operations Clarifications," dated August 11, 2008 (ADAMS Accession No. ML082200528).
- 57.
Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of National Fire Protection Association Standard 805 Frequently Asked Question 08-0046: Incipient Fire Detection Systems," dated December 1, 2009 (ADAMS Accession No. ML093220426).
- 58.
Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Closure of National Fire Protection Association 805 Frequently Asked Question 08-0048 Revised Fire Ignition Frequencies," dated September 1, 2009 (ADAMS Accession No. ML092190457).
- 59.
Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of National Fire Protection Association Frequently Asked 08-0054 on Demonstrating Compliance with Chapter 4 of National Fire Protection Association 805,"
dated February 17,2011 (ADAMS Accession No. ML110140183).
- 60.
Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of National Fire Protection Association 805 Frequently Asked Question 09-0056 on Radioactive Release Transition," dated January 14, 2011 (ADAMS Accession No. ML102920405).
- 61.
Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of National Fire Protection Association Standard 805 Frequently Asked Question 10-0059: National Fire Protection 805 Monitoring Program," dated March 19, 2012 (ADAMS Accession No. ML120750108).
- 164 -
- 62.
Nichols, TC, South Carolina Electric and Gas Company, Letter to Denton, Harold, USN RC, Forwards lists of Facility Compliance with 10 CFR 20, 10 CFR 50, and 10 CFR 100, November 14, 1990, (ADAMS Accession No. 8011210503).
- 63.
Marion, Alexander, Nuclear Energy Institute, letter dated June 17, 2003, to John Hannon, U.S. Nuclear Regulatory Commission, transmitting Revision 0 of NEI 02-03, "Guidance for Performing a Regulatory Review of Proposed Changes to the Approved Fire Protection Program," June 2003 (ADAMS Accession No. ML031780500).
- 64.
National Fire Protection Association Standard 72, (NFPA 72), "National Fire Alarm Code",
Quincy, Massachusetts.
- 65.
Electric Power Research Institute (EPRI) Technical Report TR-1006756, "Fire Protection Equipment Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features," Final Report, Palo Alto, CA, Final Report July 2003.
- 66.
National Fire Protection Association Standard 76 (2009), (NFPA 76-2009), "Standard for the Fire Protection of Telecommunications Facilities" Quincy, Massachusetts.
- 67.
National Fire Protection Association Standard 101, (NFPA 101 ), "Life Safety Code" Quincy, Massachusetts.
- 68.
National Fire Protection Association Standard 256, (NFPA 256), "Standard Methods of Fire Test of Roof Coverings," Quincy, Massachusetts.
- 69.
National Fire Protection Association Standard 80, (NFPA 80), "Standard for Fire Doors and Fire Windows," Quincy, Massachusetts.
- 70.
National Fire Protection Association Standard 90A, (NFPA 90A), "Standard for Installation of Air Conditioning and Ventilating Systems," Quincy, Massachusetts.
- 71.
Institute of Electrical and Electronics Engineers, IEEE-383: "Standard for Qualifying Class 1 E Electric Cables and Field Splices for Nuclear Power Generating Stations."
- 72.
Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of National Fire Protection Association Frequently Asked Question 06-0022 on Electrical Cable Flame Propagation Tests/' dated May 5, 2009 (ADAMS Accession No. ML091240278).
- 73.
National Fire Protection Association Standard 24 (1973), (NFPA 24-1973), "Standard for the Installation of Private Fire Service Mains and Their Appurtenances" Quincy, Massachusetts.
- 74.
Appendix A to BTP APCSB 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976" (ADAMS Accession No. ML070660458).
- 75.
National Fire Protection Association Standard 14 (1974), (NFPA 14-1974), "Standard for the Installation of Standpipe and Hose Systems" Quincy, Massachusetts.
- 165 -
- 76.
National Fire Protection Association Standard 72E (1978), (NFPA 72E-1978), "Standard for Automatic Fire Detectors" Quincy, Massachusetts.
- 77.
National Fire Protection Association Standard 72 (1990), (NFPA 72-1990), "National Fire Alarm Code", Quincy, Massachusetts.*
.78.
U.S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Chapter 9.5.1, "Fire Protection Program," Revision 3, July 1981, (ADAMS Accession No. ML052350030).
- 79.
EPRI 1016735, Fire PRA Methods Enhancements, December 2008, (ADAMS Accession No. ML090290195).
- 80.
Not Used.
- 81.
U.S. Nuclear Regulatory Commission, NUREG-1934, "Nuclear Power Plant Fire Modeling Analysis Guidelines (NPP FIRE MAG)," November 2012 (ADAMS Accession No. ML12314A165).
- 82.
NEI 00-02, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance,"
Revision A3, Nuclear Energy Institute (NEI), Washington, DC, March 20, 2000.
- 83.
Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 1, U. S.
Nuclear Regulatory Commission, Washington, DC, January 2007, (ADAMS Accession No. ML070240001 ).
- 84.
U.S. Nuclear Regulatory Commission, "Record of Review, Virgil C. Summer Nuclear Station, Unit 1, LAR Attachment U-Table U-1 Internal Events PRA Peer Review-Facts and Observations (F&Os)," and "Record of Review, Virgil C. Summer Nuclear Station, Unit 1, LAR Attachment V-Tables V-1 and V-2 Fire PRA Peer Review-Facts and Observations (F&Os)," November 7, 2014 (ADAMS Accession Nos. ML14311A128 and ML14311A140, respectively).
- 85.
NEI 07-12, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines, Revision 1, Nuclear Energy Institute (NEI), June 2010, Washington, DC.
- 86.
- Hamzehee, Hossein, G., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of Fire Probabilistic Risk Assessment Frequently Asked Question 13-0005 on Cable Fires Special Cases: Self-Ignited and Caused by Welding and Cutting," dated December 3, 2013, (ADAMS Accession No. ML133198181).
- 87.
Hamzehee, Hossein, G., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of Fire Probabilistic Risk Assessment Frequently Asked Question 13-0006 on Modeling Junction Box Scenarios in a Fire PRA," dated December 12, 2013, (ADAMS Accession No. ML 1333a8213).
- 166 -
- 88.
Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of National Fire Protection Association Frequently Asked Question 12-0064 on Hot Work/Transient Fire Frequency Influence Factors," dated January 17, 2013 (ADAMS Accession No. ML12346A488).
- 89.
Westinghouse Electric Company, LLC, "Model for Failure of RCP Seals Given Loss of Seal Cooling in CE NSSS Plants," WCAP-16175-P, Revision 0, dated January 2004 (ADAMS Accession No. ML040340226).
- 90.
Nieh, Ho K., U.S. Nuclear Regulatory Commission, letter to Gordon Bischoff, Westinghouse Electric Company, "Final Safety Evaluation for Pressurized Water Reactor Owners Group (PWROG) Topical Report WCAP-16175-P, Revision 0, (CE NPSD-1199, Revision 1) "Model for Failure of RCP Seals Given Loss of Seal Cooling in CE NSSS Plants" (TAC No. MB5803)," dated February 12, 2007 (ADAMS Accession No. ML070240429).
- 91.
Evaluation of the Computer Fire Model DETACT-QS, SFPE Engineering Guide, Society of Fire Protection Engineers, Bethesda, MD, 2002.
- 92.
Memo from Patrick Hiland, NRC, to Brian Sheron, NRC, "Safety/Risk Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants." Washington, D.C. dated September 2, 2010 (ADAMS Accession No. ML 10020598).
- 93.
GL 86-10, Supplement 1, "Fire Endurance Test Acceptance Criteria for Fire Barrier Systems used to Separate Redundant Safe Shutdown Trains within the Same Fire Area."
- 94.
SFPE Handbook of Fire Protection Engineering, 4th Edition, DiNenno P.J., Editor in-Chief, National Fire Protection Association, Quincy, Massachusetts, 2008.
- 95.
Heskestad, G., "Fire Plumes, Flame Height, and Air Entrainment," Chapter 2-1, SFPE Handbook of Fire Protection Engineering, 4th Edition, National Fire Protection Association, Quincy, Massachusetts, 2008.
- 96.
Seyler, C. L., "Fire Hazard Calculations for Large, Open Hydrocarbon Fires," Chapter 3-10, SFPE Handbook of Fire Protection Engineering, 4th Edition, P. J. Di Nenno, Editor-in-Chief, National Fire Protection Association, Quincy, Massachusetts, 2008.
- 97.
Budnick, E.K., D.D. Evans, and H.E. Nelson, "Simplified Fire Growth Calculations,"
Section 3, Chapter 9, NFPA Fire Protection Handbook, 19th Edition, A.E. Cote, Editor-in-Chief, National Fire Protection Association, Quincy, Massachusetts, 2003.
- 98.
Peacock, RD., McGrattan, K., Klein, B., Jones, W.W., and Reneke, P.A., "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 (NIST), Gaithersburg, Maryland, December 2008.
- 167 -
- 99.
Johnson, AR., U.S. Nuclear Regulatory Commission, letter to Taylor, G.J., South Carolina Electric and Gas Company, Deviation from 10 CFR Part.50, Appendix R, Section 111.G. Fire Protection of Safe Shutdown Capability for Virgil C. Summer Nuclear Station (TAC No. M97337), dated October 19, 1997 (ADAMS Legacy Library Accession No.
9710280356).
100.
Archie, Jeffrey, B., South Carolina Electric and Gas Company, letter to U.S. Nuclear Regulatory Commission, Virgil C. Summer Nuclear Station, Docket No. 50/395, Operating License No. NPF-12, 60-Day Response to NRC Generic Letter 2006-03: Potentially Nonconforming Hemyc and MT Fire Barrier Configurations, dated June 5, 2006 (ADAMS Accession No. ML061590311 ).
Principal Contributors:
Jay Robinson, NRR Harold Barrett, NRR Brian Metzger, NRR Alayna Pearson, NRR Gary Cooper, NRR JS Hyslop, NRR Steven Garry, NRR Karl Bohlander, PNNL Bill Ivans, PNNL Marc Janssens, CNWRA Debashis Basu, CNWRA Robert Fosdick, CNWRA Date: February 11, 2015 Attachments:
A.
Table 3.8 V&V Basis for Fire Modeling Correlations Used at VCSNS B.
Table 3.8 V&V Basis for Fire Model Calculations of Other Models Used at VCSNS C.
Abbreviations and Acronyms
Attachment A: Table 3.8-1, V&V Basis for Fire Modeling Correlations Used at VCSNS Correlation Application at V&V Basis NRC Staff Evaluation of Acceptability VCSNS Plume Centerline The Plume NUREG-1805, Licensee provided verification of the implementation of this Temperature Centerline I
Chapter 9, 2004 correlation in the VCSNS Fire Modeling Database based Temperature (Reference 47) on a comparison with NUREG-1805 (Response to FM RAI (Method of correlation was 03.a)
Heskestad) used to calculate the NUREG-1824, vertical separation Volume 3, 2007 The correlation is validated in NUREG-1824 and an distance, based on (Reference 48) authoritative publication of the SFPE Handbook temperature, to a target in order to SFPE Handbook, Licensee stated that in most cases, the correlation has determine the 4th Edition, been applied within the validated range reported in vertical extent of the Chapter 2-1, NUREG-1824. Licensee provided justification for cases ZOI.
Heskestad, 2008 where the correlation was used outside the validated range (Reference 95) reported in NUREG-1824 (Response to FM RAI 03.d).
Based on these observations, the NRC staff concludes that the use of this correlation in the VCSNS application acceptable.
NUREG-1805, Licensee provided verification of the implementation of this The Radiant Heat Chapter 5, 2004 correlation in the VCSNS Fire Modeling Database based Flux (Point Source (Reference 4 7) on a comparison with NUREG-1805 (Response to FM RAI Method) correlation 03.a) was used to NUREG-1824, determine the Radiant Heat Flux horizontal Volume 3, 2007 The correlation is validated in NUREG-1824 and an (Point Source separation distance, (Reference 48) authoritative publication of the SFPE Handbook Method) based on heat flux, to a target in order to SFPE Handbook, Licensee stated that in most cases, the correlation has 4th Edition, been applied within the validated range reported in determine the horizontal extent of Chapter 3-10, NUREG-1824. Licensee provided justification for cases the ZOI.
Seyler, C., 2008 where the correlation was used outside the validated range (Reference 96) reported in NUREG-1824 (Response to RAI 03.d).
Attachment Attachment A: Table 3.8-1, V&V Basis for Fire Modeling Correlations Used at VCSNS Correlation Application at V&V Basis NRC Staff Evaluation of Acceptability VCSNS Based on these observations, the NRC staff found the use of this correlation in the VCSNS application acceptable.
NUREG-1805, Chapter 10, 2004 (Reference 47)
NFPA Handbook, The Sprinkler 19th Edition, Activation Model Chapter 3-9, was used to Budnick, E.,
estimate sprinkler Evans, D., and
- The correlation is validated in an authoritative publication.
Sprinkler actuation timing Nelson, H., 2003 Activation Model based on ceiling jet (Reference 97)
Based on these observations, the NRC staff found the use of this temperature, correlation in the VCSNS application acceptable.
velocity, and SFPE Engineering thermal response of Guide on the sprinkler head.
Evaluation of the Computer Fire Model DETACT-QS, 2002 (Reference 91)
Attachment B: Table 3.8-2, V&V Basis for Other Models Used at VCSNS Model Application at V&V Basis NRC Staff Evaluation of Acceptability VCSNS CFAST (Version CFAST (Version 6)
- The model is validated in NUREG-1824 and an
- 6) was used to Volume 5, 2007 authoritative publication of NIST.
calculate hot gas (Reference 48) layer height and Licensee stated that in most cases, the correlation has temperature for NIST Special been applied within the validated range reported in various Publication 1086, NUREG-1824. Licensee provided justification for cases compartments. It 2008 where the correlation was used outside the validated range was also used to (Reference 98) reported in NUREG-1824 (Response to RAI 03.d).
calculate abandonment time Based on these observations, the NRC staff found the use of this for the main control correlation in the VCSNS application acceptable.
room.
AC ADAMS AFW AHJ ANS APCSB ASD ASME BTP BWR CA CAFP CCDP CCF ccw CDF CF AST CFR CFWC CHRISTI FIRE CNWRA CPT CREP CRS CT DC DID DID RA ECA Epsilon (e:)
EEEE EP EPRI ERFBS ERO ESW F&O F&S FAQ FDS FDT FHRA FIVE FLASH-CAT FM FMDB FPE Attachment C: Abbreviations and Acronyms alternating current Agencywide Documents Access and Management System auxiliary feedwater authority having jurisdiction American Nuclear Society Auxiliary and Power Conversion Systems Branch Aspirating Smoke Detectors American Society of Mechanical Engineers Branch Technical Position boiling-water reactor complies by alternative conditional abandonment failure probability conditional core damage probability common-cause failure component cooling water core damage frequency consolidated model of fire and smoke transport Code of Federal Regulations caused by welding and cutting Cable Heat Release, Ignition, and Spread in Tray Installations During Fire center for nuclear waste regulatory analysis control power transformer control room evacuation panel control room supervisor current transformer direct current defense-in-depth defense-in-depth recovery action equipment cabinet area Non-zero but below truncation limit existing engineering equivalency evaluation Emergency Plan Electric Power Research Institute electrical raceway fire barrier system emergency response organization essential service water facts and observations findings and suggestions frequently asked question fire dynamics simulator fire dynamics tool Fire Human Reliability Analysis Fire Induced Vulnerability Evaluation Methodology Flame Spread over Horizontal Cable Trays fire modeling fire modeling database fire protection engineering
FPEEE FPER FPP FPRA FR FRE FSAR ft GDC GL gpm.
HOPE HEAF HEP HFE HGL HRA HRE HRR IN in.
IEEE ISLOCA KSF kW LAN LAR LERF LFS LOCA MAAP MCA MCB MCR MEFS min MSO NEIL NEI NIST NFPA No.
NPO NPP NRC NRR NSCA NSP Fire Protection Engineering Equivalency Evaluation Fire Protection Evaluation Report fire protection program fire probabilistic risk assessment Federal Register fire risk evaluation final safety analysis report foot general design criteria generic letter gallons per minute high-density polyethylene high-energy arching fault human error probability human failure event hot gas layer human reliability analysis high(er) risk evolution heat release rate information notice inches Institute of Electrical and Electronics Engineers interfacing systems loss of coolant accident key safety function kilowatt licensing action number license amendment request large early release frequency limiting fire scenario loss-of-coolant accident Modular Accident Analysis Program multi-compartment analysis main control board main control room maximum expected fire scenario minute(s) multiple spurious operation Nuclear Electric Insurance Limited Nuclear Energy Institute National Institute of Standards and Technology National Fire Protection Association number non-power operation nuclear power plant U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation nuclear safety capability assessment non-suppression probability
NSPC ODCM OQAM OMA P&ID PACR PB PCE PCS PNNL PORV POS PRA PSA psi PWR PWROG QA RA RAI RCP RCS RES RG RHR RI RI/PB RIS RTI SE SER SFPE SG SGTR SISBO SNUPPS SR SSA SSC SSE TAWs TDAFW TR TS UHS UL V&V VAC nuclear safety performance criteria offsite dose calculation manual Operating Quality Assurance Manual operator manual action piping and instrumentation drawing prior approval clarification request performance-based plant change evaluation primary control station pacific northwest national laboratory power-operated relief valve plant operational state probabilistic risk assessment probabilistic safety assessment pounds per square inch pressurized-water reactor PWR Owner's Group quality assurance recovery action request for additional information reactor coolant pump reactor coolant system Office of Nuclear Regulatory Research Regulatory Guide residual heat removal risk-informed.
risk-informed, performance-based regulatory issue summary response time index safety evaluation safety evaluation report Society of Fire Protection Engineers steam generator steam generator tube rupture self-induced station blackout Standardized Nuclear Unit Power Plant System supporting requirements safe shutdown analysis structures, systems, and components safe shutdown earthquake Transient Analysis Worksheets turbine-driving auxiliary feedwater technical/topical report technical specifications ultimate heat sink Underwriters Laboratory verification and validation volts alternating current
VEWFDS VFDR WOG YD yr ZOI very early warning fire detectors variance from deterministic requirements Westinghouse Owners Group yard year zone* of influence
A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's Biweekly Federal Register notice.
Docket No. 50-395
Enclosures:
- 1. Amendment No. 199to NPF-12
- 2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:
PUBLIC LPL2-1 R/F RidsAcrs_AcnwMailCTR Resource RidsNrrDorlDpr Resource RidsNrrDraAfpb Resource RidsNrrDraAhpb Resource RidsNrrDraApla Resource ADAMS A ccess1on N ML14287A289
- 0.
OFFICE NRR/DORL/805/PM NRR/DORULPL2-1 /PM NAME SWall SWilliams DATE 12/12/14 11/20/14 OFFICE NRR/DRA/AHPB/BC NRR/DSS/STSB/BC NAME UShoop*
RElliott DATE 11/06/14 11/24/14.
Sincerely,
/RA/
Shawn Williams, Senior Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation RidsNrrDssStsb Resource RidsNrrLASFigueroa Resource RidsRgn2MailCenter Resource RidsNrrPMSummer Resource SWilliams, NRR
- b >y memo d t d 11/6/14 ae NRR/DORL/LPL2-1/LA NRR/DRA/APLA/BC NRR/DRA/AFPB/BC SFigueroa HHa.mzehee*
AKlein*
11/20/14 11/06/14 11/06/14 OGC NRR/DORL/LPL2-1 /BC NRR/DORL/LPL2-1 /PM MYoung/NLO RPascarelli SWilliams 01 /30/15 02/11/15 02/11/15 Official Record Copy