ML11332A077: Difference between revisions

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NEI developed NEI 04-02 to assist licensees in adopting NFPA 805 and making the transition from their current fire protection licensing basis to one based on NFPA 805.The NRC issued Regulatory Guide (RG) 1.205, Risk-Informed, Performance-Based Fire Protection for Existing Light Water Nuclear Power Plants, which endorses NEI 04-02, with exceptions, in December 2009.1 A depiction of the primary document relationships is shown in Figure 1-1: 1 Where referred to in this document NEI 04-02 is Revision 2 and RG 1.205 is Revision 1.Introduction Page 1 Introduction Page 1 Aý8 RC-11-0149 Section 1.0 Incorporation by E a-0)Perfermanee Based StAndar~dfor FIRfor PI LqtWa pa-n-Incorporation by Reference z Z 10 CFR 50.48(c)National Fire Protection Association
NEI developed NEI 04-02 to assist licensees in adopting NFPA 805 and making the transition from their current fire protection licensing basis to one based on NFPA 805.The NRC issued Regulatory Guide (RG) 1.205, Risk-Informed, Performance-Based Fire Protection for Existing Light Water Nuclear Power Plants, which endorses NEI 04-02, with exceptions, in December 2009.1 A depiction of the primary document relationships is shown in Figure 1-1: 1 Where referred to in this document NEI 04-02 is Revision 2 and RG 1.205 is Revision 1.Introduction Page 1 Introduction Page 1 Aý8 RC-11-0149 Section 1.0 Incorporation by E a-0)Perfermanee Based StAndar~dfor FIRfor PI LqtWa pa-n-Incorporation by Reference z Z 10 CFR 50.48(c)National Fire Protection Association
_- Standard NFPA 805 EEEFMý'CD NEI 04-02 GUIDANCE FOR IMPLEMENTING A RI-PB FP PROGRAM UNDER 10 CFR 50.48(c)Endorsement RG 1.205 RI-PB FP FOR EXISTING LIGHT-WATER NUCLEAR POWER PLANTS Figure 1-1 NFPA 805 Transition  
_- Standard NFPA 805 EEEFMý'CD NEI 04-02 GUIDANCE FOR IMPLEMENTING A RI-PB FP PROGRAM UNDER 10 CFR 50.48(c)Endorsement RG 1.205 RI-PB FP FOR EXISTING LIGHT-WATER NUCLEAR POWER PLANTS Figure 1-1 NFPA 805 Transition  
-Implementation Requirements/Guidance 1.1.2 Transition to 10 CFR 50.48(c)1.1.2.1 Start of Transition In October 2006, VCSNS decided to transition the fire protection licensing basis to the RI-PB alternative in 10 CFR 50.48(c).
-Implementation Requirements/Guidance
 
====1.1.2 Transition====
 
to 10 CFR 50.48(c)1.1.2.1 Start of Transition In October 2006, VCSNS decided to transition the fire protection licensing basis to the RI-PB alternative in 10 CFR 50.48(c).
SCE&G submitted a letter of intent to the NRC on October 19, 2006 (ML062990453) for VCSNS to adopt NFPA 805 in accordance with 10 CFR 50.48(c).By letter dated January 19, 2007 (ML063520409), the NRC granted a three year enforcement discretion period from October 19, 2006 to October 19, 2009. In accordance with NRC Enforcement Policy, the enforcement discretion period will continue until the NRC approval of the license amendment request (LAR) is completed.
SCE&G submitted a letter of intent to the NRC on October 19, 2006 (ML062990453) for VCSNS to adopt NFPA 805 in accordance with 10 CFR 50.48(c).By letter dated January 19, 2007 (ML063520409), the NRC granted a three year enforcement discretion period from October 19, 2006 to October 19, 2009. In accordance with NRC Enforcement Policy, the enforcement discretion period will continue until the NRC approval of the license amendment request (LAR) is completed.
On July 16, 2009 SCE&G submitted a request in accordance with COMSECY-08-022 to extend the enforcement discretion to six months past the date of the safety evaluation approving the second pilot plant LAR review. In a letter (ML092920297) dated October 19, 2009, the NRC granted this enforcement discretion extension request.The NRC expected approximately 23 LARs by the end of June 2011. As a result, the Commission worked with industry to develop and create a staggered LAR submittal schedule.
On July 16, 2009 SCE&G submitted a request in accordance with COMSECY-08-022 to extend the enforcement discretion to six months past the date of the safety evaluation approving the second pilot plant LAR review. In a letter (ML092920297) dated October 19, 2009, the NRC granted this enforcement discretion extension request.The NRC expected approximately 23 LARs by the end of June 2011. As a result, the Commission worked with industry to develop and create a staggered LAR submittal schedule.
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Under the FAQ Process, transition issues are submitted to the NEI NFPA 805 Task Force for review, and subsequently presented to the NRC during public FAQ meetings.Once the NEI NFPA 805 Task Force and NRC reach agreement, the NRC issues a closure memorandum to indicate that the FAQ is acceptable.
Under the FAQ Process, transition issues are submitted to the NEI NFPA 805 Task Force for review, and subsequently presented to the NRC during public FAQ meetings.Once the NEI NFPA 805 Task Force and NRC reach agreement, the NRC issues a closure memorandum to indicate that the FAQ is acceptable.
NEI 04-02 will be revised to incorporate the approved FAQs. This is an on-going revision process that will continue through the transition of NFPA 805 non-pilot plants. Final closure of the FAQs will occur when future revisions of RG 1.205, endorsing the related revisions of NEI 04-02, are approved by the NRC. It is expected that additional FAQs will be written and existing FAQs will be revised as plants continue NFPA 805 transition after NRC approval of the Pilot Plant Safety Evaluations.
NEI 04-02 will be revised to incorporate the approved FAQs. This is an on-going revision process that will continue through the transition of NFPA 805 non-pilot plants. Final closure of the FAQs will occur when future revisions of RG 1.205, endorsing the related revisions of NEI 04-02, are approved by the NRC. It is expected that additional FAQs will be written and existing FAQs will be revised as plants continue NFPA 805 transition after NRC approval of the Pilot Plant Safety Evaluations.
Attachment H contains the list of approved FAQs not yet incorporated into the endorsed revision of NEI 04-02. These FAQs have been reviewed and/or used to clarify the guidance in RG 1.205, NEI 04-02, and the requirements of NFPA 805 and in the preparation of this LAR.Transition Process Page 10 Transition Process Page 10 A -RC-11-0149 Section 4.0 4.0 COMPLIANCE WITH NFPA 805 REQUIREMENTS 4.1 Fundamental Fire Protection Program and Design Elements The Fundamental Fire Protection Program and Design Elements are established in Chapter 3 of NFPA 805. Section 4.3.1 of NEI 04-02 provides an industry guideline and systematic process for determining the extent to which the pre-transition licensing basis and plant configuration meets these criteria and for identifying the fire protection program changes that would be necessary for compliance with NFPA 805. SCE&G has developed the Fire Protection Program compliance review with the basic guidance, process and criteria promulgated within these documents.
Attachment H contains the list of approved FAQs not yet incorporated into the endorsed revision of NEI 04-02. These FAQs have been reviewed and/or used to clarify the guidance in RG 1.205, NEI 04-02, and the requirements of NFPA 805 and in the preparation of this LAR.Transition Process Page 10 Transition Process Page 10 A -RC-11-0149 Section 4.0 4.0 COMPLIANCE WITH NFPA 805 REQUIREMENTS
4.1.1 Overview of Evaluation Process The comparison of the VCSNS Fire Protection Program to the requirements of NFPA 805 was performed and documented in design calculations that were developed in accordance with the configuration control and quality inherent in SCE&G Design Engineering and as indicated in Section 2.7 of NFPA 805. The NFPA 805 Chapter 3 Code Compliance Document used the guidance contained in NEI 04-02, Section 4.3.1 and Appendix B-1 to develop the documentation package(s) (Reference Figure 4-1).In addition to NFPA 805, Chapter 3 [10 CFR 50.48(c)], applicable codes and/or standards (codes of record) that may have been utilized as design input and are essential to the functional performance of the system and/or feature being reviewed has been incorporated into this overall compliance assessment through the use of individual design calculations that represent NFPA codes that are applicable to VCSNS.The application of these NFPA codes, relate to fire protection systems, structures and features that are considered part of the "power block", as described in NFPA 805 Section 1.6.46, and are "required" systems or features to support the results of the analysis as described in NFPA 805. This includes fire protection systems, structures and features for those plant areas that may be credited to support the nuclear safety performance criteria described in NFPA 805 Section 1.5.1, Nuclear Safety Performance Criteria, or performance based evaluations (see Section 4.8.2 of the Transition Report).Each section and subsection of NFPA 805 Chapter 3 (Table B-i) was reviewed against the current station fire protection program. In some cases multiple compliance statements may have been assigned to a specific NFPA 805 Chapter 3 element.Where this is the case, each compliance basis statement clearly references the corresponding requirement of NFPA 805 Chapter 3. When other NFPA Codes are referenced out of NFPA 805 Chapter 3, the analysis and compliance will be performed and evaluated in a similar manner, as applicable, to the design and features at the station, subject to review and approval by the SCE&G qualified fire protection engineer.Compliance with NFPA 805 Requirements Page 11 Compliance with NFPA 805 Requirements Page I11 RC-11-0149 Section 4.0 Existing~~Secio 4.0-t. ie rtctFm arida~ DeslignrEle~me Ene H In the Compliance Basis H In Reference Document....ld provide toerle ,Field provide:
 
===4.1 Fundamental===
 
Fire Protection Program and Design Elements The Fundamental Fire Protection Program and Design Elements are established in Chapter 3 of NFPA 805. Section 4.3.1 of NEI 04-02 provides an industry guideline and systematic process for determining the extent to which the pre-transition licensing basis and plant configuration meets these criteria and for identifying the fire protection program changes that would be necessary for compliance with NFPA 805. SCE&G has developed the Fire Protection Program compliance review with the basic guidance, process and criteria promulgated within these documents.
 
====4.1.1 Overview====
of Evaluation Process The comparison of the VCSNS Fire Protection Program to the requirements of NFPA 805 was performed and documented in design calculations that were developed in accordance with the configuration control and quality inherent in SCE&G Design Engineering and as indicated in Section 2.7 of NFPA 805. The NFPA 805 Chapter 3 Code Compliance Document used the guidance contained in NEI 04-02, Section 4.3.1 and Appendix B-1 to develop the documentation package(s) (Reference Figure 4-1).In addition to NFPA 805, Chapter 3 [10 CFR 50.48(c)], applicable codes and/or standards (codes of record) that may have been utilized as design input and are essential to the functional performance of the system and/or feature being reviewed has been incorporated into this overall compliance assessment through the use of individual design calculations that represent NFPA codes that are applicable to VCSNS.The application of these NFPA codes, relate to fire protection systems, structures and features that are considered part of the "power block", as described in NFPA 805 Section 1.6.46, and are "required" systems or features to support the results of the analysis as described in NFPA 805. This includes fire protection systems, structures and features for those plant areas that may be credited to support the nuclear safety performance criteria described in NFPA 805 Section 1.5.1, Nuclear Safety Performance Criteria, or performance based evaluations (see Section 4.8.2 of the Transition Report).Each section and subsection of NFPA 805 Chapter 3 (Table B-i) was reviewed against the current station fire protection program. In some cases multiple compliance statements may have been assigned to a specific NFPA 805 Chapter 3 element.Where this is the case, each compliance basis statement clearly references the corresponding requirement of NFPA 805 Chapter 3. When other NFPA Codes are referenced out of NFPA 805 Chapter 3, the analysis and compliance will be performed and evaluated in a similar manner, as applicable, to the design and features at the station, subject to review and approval by the SCE&G qualified fire protection engineer.Compliance with NFPA 805 Requirements Page 11 Compliance with NFPA 805 Requirements Page I11 RC-11-0149 Section 4.0 Existing~~Secio 4.0-t. ie rtctFm arida~ DeslignrEle~me Ene H In the Compliance Basis H In Reference Document....ld provide toerle ,Field provide:
* Document References that'Compies
* Document References that'Compies
* No additional demonstrate compliance Document any Open in Compliance Satement clarification Items found duing Review fildieldesoy Eater 'CorEnter u In t Coropliance Basis ield In Refence Document FlPet prvdide: Nis.NoNg 'Coesples nirth Fild prOVide
* No additional demonstrate compliance Document any Open in Compliance Satement clarification Items found duing Review fildieldesoy Eater 'CorEnter u In t Coropliance Basis ield In Refence Document FlPet prvdide: Nis.NoNg 'Coesples nirth Fild prOVide
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Complies through the use of Fire Protection Engineering Equivalency Evaluations (FPEEE) which are valid and of appropriate quality. VCSNS requests NRC review/approval of those Engineering Evaluations listed in Section 4.1.2.3 (Table 4-1) of the Transition Report and included in Attachment L." Complies by Previous NRC Approval (CNRC) -The existing FPP elements specified in NFPA 805 Chapter 3 requirements are not in strict compliance, however, previous NRC approval of the configuration exists. An NRC approved alternative or deviation to NFPA 805 Chapter 3, would supplant the specific requirement of NFPA 805 Chapter 3. Where credited, these prior approvals have been incorporated into an FPEEE, and included in Attachment K and Attachment L." No Review Required (NRR) -The existing Chapter 3 elements are not based on the requirements and/or are not applicable to elements of the VCSNS Fire Protection Program.In addition to these compliance statements, the following approaches were implemented: " An "open item" in any category of the B-1 Table means there are action items to be completed during implementation prior to full transition of the Fire Protection Program. These open items are identified in Attachment S, Table S-2." The use of FPEEEs in the compliance review process is identified by the "CE" designation.
Complies through the use of Fire Protection Engineering Equivalency Evaluations (FPEEE) which are valid and of appropriate quality. VCSNS requests NRC review/approval of those Engineering Evaluations listed in Section 4.1.2.3 (Table 4-1) of the Transition Report and included in Attachment L." Complies by Previous NRC Approval (CNRC) -The existing FPP elements specified in NFPA 805 Chapter 3 requirements are not in strict compliance, however, previous NRC approval of the configuration exists. An NRC approved alternative or deviation to NFPA 805 Chapter 3, would supplant the specific requirement of NFPA 805 Chapter 3. Where credited, these prior approvals have been incorporated into an FPEEE, and included in Attachment K and Attachment L." No Review Required (NRR) -The existing Chapter 3 elements are not based on the requirements and/or are not applicable to elements of the VCSNS Fire Protection Program.In addition to these compliance statements, the following approaches were implemented: " An "open item" in any category of the B-1 Table means there are action items to be completed during implementation prior to full transition of the Fire Protection Program. These open items are identified in Attachment S, Table S-2." The use of FPEEEs in the compliance review process is identified by the "CE" designation.
These are utilized to evaluate the requirement and field conditions Compliance with NFPA 805 Requirements Page 13  
These are utilized to evaluate the requirement and field conditions Compliance with NFPA 805 Requirements Page 13  
° RC-11-0149 Section 4.0 and determine the level of compliance and if the element or feature is"Equivalent" or "Adequate for the Hazard." They may also be utilized to assist in the clarification of requirements, past approvals (CNRC) and field conditions that involve complex systems, features or elements that need further understanding within the FPP. The "CE" may be a self-approval, but available for review (see Section 4.2.2 of the Transition Report).The VCSNS process to "self-approve" selected Chapter 3 (Sections 3.8 through 3.11 of NFPA 805) elements and or submit them for NRC review and approval (Sections 3.1 through 3.7 of NFPA 805) is in accordance with the guidance provided in FAQ 06-0008 and RG 1.205 Revision 1.Note: Specific references to VCSNS controlled documents used in this transition report are for reference use only; similar documents, updated revisions or other forms of media may be used to manage this type of information in the future.4.1.2.2 NFPA 805 Chapter 3: "Previous NRC Approval (CNRC)" NFPA 805 Section 3.1 states in part, "Previously approved alternatives from the fundamental protection program attributes of this chapter by the AHJ take precedence over the requirements contained herein." In some cases the prior NRC approval may be unclear for an NFPA 805 Chapter 3 program attribute, to support future clarity for NRC inspections.
° RC-11-0149 Section 4.0 and determine the level of compliance and if the element or feature is"Equivalent" or "Adequate for the Hazard." They may also be utilized to assist in the clarification of requirements, past approvals (CNRC) and field conditions that involve complex systems, features or elements that need further understanding within the FPP. The "CE" may be a self-approval, but available for review (see Section 4.2.2 of the Transition Report).The VCSNS process to "self-approve" selected Chapter 3 (Sections  
 
===3.8 through===
3.11 of NFPA 805) elements and or submit them for NRC review and approval (Sections  
 
===3.1 through===
3.7 of NFPA 805) is in accordance with the guidance provided in FAQ 06-0008 and RG 1.205 Revision 1.Note: Specific references to VCSNS controlled documents used in this transition report are for reference use only; similar documents, updated revisions or other forms of media may be used to manage this type of information in the future.4.1.2.2 NFPA 805 Chapter 3: "Previous NRC Approval (CNRC)" NFPA 805 Section 3.1 states in part, "Previously approved alternatives from the fundamental protection program attributes of this chapter by the AHJ take precedence over the requirements contained herein." In some cases the prior NRC approval may be unclear for an NFPA 805 Chapter 3 program attribute, to support future clarity for NRC inspections.
In other cases, the requirement has changed from the originally licensed Fire Protection Program attribute.
In other cases, the requirement has changed from the originally licensed Fire Protection Program attribute.
VCSNS requests that the NRC concur with their finding of prior approval and acceptability for the following sections of NFPA 805 Chapter 3 designated as (CNRC).0 None.4.1.2.3 NFPA 805 Chapter 3: "Compliance Alternatives (CA) Not Previously Approved by NRC" The sections of NFPA 805 Chapter 3 may not have previous NRC approval of an alternate approach, methods and/or condition which VCSNS considers to be minor variations to, and are equivalent to the NFPA 805 requirements.
VCSNS requests that the NRC concur with their finding of prior approval and acceptability for the following sections of NFPA 805 Chapter 3 designated as (CNRC).0 None.4.1.2.3 NFPA 805 Chapter 3: "Compliance Alternatives (CA) Not Previously Approved by NRC" The sections of NFPA 805 Chapter 3 may not have previous NRC approval of an alternate approach, methods and/or condition which VCSNS considers to be minor variations to, and are equivalent to the NFPA 805 requirements.
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Clarification and approval is requested for the existing fire detection layout of devices that are in accordance with NFPA 72E-1 978 code of record.Compliance with NFPA 805 Requirements Page 15 Compliance with NFPA 805 Requirements Page 15 A 'a RC-11-0149 Section 4.0 4.1.3 Definition of Power Block and Plant Where used in NFPA 805 Chapter 3 the terms "Power Block" and "Plant" refer to structures that have equipment required for nuclear plant operations, such as Containment, Auxiliary Building, Service Building, Control Building, Fuel Building, Radioactive Waste, Water Treatment, Turbine Building, and intake structures or structures that are identified in the facility's pre-transition licensing basis.VCSNS reviewed the structures in the Owner Controlled Area to determine those that contain equipment that is required to meet the nuclear safety performance criteria described in Section 1.5 of NFPA 805 and are required for nuclear plant operations.
Clarification and approval is requested for the existing fire detection layout of devices that are in accordance with NFPA 72E-1 978 code of record.Compliance with NFPA 805 Requirements Page 15 Compliance with NFPA 805 Requirements Page 15 A 'a RC-11-0149 Section 4.0 4.1.3 Definition of Power Block and Plant Where used in NFPA 805 Chapter 3 the terms "Power Block" and "Plant" refer to structures that have equipment required for nuclear plant operations, such as Containment, Auxiliary Building, Service Building, Control Building, Fuel Building, Radioactive Waste, Water Treatment, Turbine Building, and intake structures or structures that are identified in the facility's pre-transition licensing basis.VCSNS reviewed the structures in the Owner Controlled Area to determine those that contain equipment that is required to meet the nuclear safety performance criteria described in Section 1.5 of NFPA 805 and are required for nuclear plant operations.
Note: Structures meeting the radioactive release criteria described in Section 1.5 of NFPA 805, but not required for nuclear plant operations, are separately screened and included in the radioactive release review as discussed in Section 4.4 and Attachment E of the Transition Report.These structures are listed in Attachment I and represent the "power block" and the"plant".4.2 Nuclear Safety Performance Criteria The Nuclear Safety Performance Criteria are established in Section 1.5 of NFPA 805.In addition, Chapter 4 of NFPA 805 provides the methodology to determine the fire protection systems and features required to achieve the performance criteria outlined in Section 1.5. Section 4.3.2 of NEI 04-02 provides a systematic process for determining the extent to which the pre-transition licensing basis meets these criteria and for identifying any necessary fire protection program changes. NEI 04-02, Appendix B-2 provides guidance on documenting the transition of Nuclear Safety Capability Assessment Methodology and the Fire Area compliance strategies.
Note: Structures meeting the radioactive release criteria described in Section 1.5 of NFPA 805, but not required for nuclear plant operations, are separately screened and included in the radioactive release review as discussed in Section 4.4 and Attachment E of the Transition Report.These structures are listed in Attachment I and represent the "power block" and the"plant".4.2 Nuclear Safety Performance Criteria The Nuclear Safety Performance Criteria are established in Section 1.5 of NFPA 805.In addition, Chapter 4 of NFPA 805 provides the methodology to determine the fire protection systems and features required to achieve the performance criteria outlined in Section 1.5. Section 4.3.2 of NEI 04-02 provides a systematic process for determining the extent to which the pre-transition licensing basis meets these criteria and for identifying any necessary fire protection program changes. NEI 04-02, Appendix B-2 provides guidance on documenting the transition of Nuclear Safety Capability Assessment Methodology and the Fire Area compliance strategies.
4.2.1 Nuclear Safety Capability Assessment Methodology The Nuclear Safety Capability Assessment (NSCA) Methodology review consists of four processes: " Establishing compliance with NFPA 805 Section 2.4.2" Establishing the Safe and Stable Conditions for the Plant" Defining Recovery Actions to be Transitioned" Evaluating Multiple Spurious Operations The methodology for demonstrating reasonable assurance that a fire during non-power operational (NPO) modes will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition is an additional requirement of 10 CFR 50.48(c) and is addressed in Section 4.3 of the Transition Report.4.2.1.1 Compliance with NFPA 805 NSCA (Section 2.4.2)Overview of Process NFPA 805 Section 2.4.2 Nuclear Safety Capability Assessment states: Compliance with NFPA 805 Requirements Page 16 Compliance with NFPA 805 Requirements Page 16 RC-11-0149 Section 4.0"The purpose of this section is to define the methodology for performing a nuclear safety capability assessment.
 
====4.2.1 Nuclear====
Safety Capability Assessment Methodology The Nuclear Safety Capability Assessment (NSCA) Methodology review consists of four processes: " Establishing compliance with NFPA 805 Section 2.4.2" Establishing the Safe and Stable Conditions for the Plant" Defining Recovery Actions to be Transitioned" Evaluating Multiple Spurious Operations The methodology for demonstrating reasonable assurance that a fire during non-power operational (NPO) modes will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition is an additional requirement of 10 CFR 50.48(c) and is addressed in Section 4.3 of the Transition Report.4.2.1.1 Compliance with NFPA 805 NSCA (Section 2.4.2)Overview of Process NFPA 805 Section 2.4.2 Nuclear Safety Capability Assessment states: Compliance with NFPA 805 Requirements Page 16 Compliance with NFPA 805 Requirements Page 16 RC-11-0149 Section 4.0"The purpose of this section is to define the methodology for performing a nuclear safety capability assessment.
The following steps shall be performed:
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" The NSCA methodology review evaluated the existing NSCA methodology against the guidance provided in NEI 00-01, Revision 1 Chapter 3, "Deterministic Methodology," as discussed in Appendix B-2 of NEI 04-02. The methodology in Figure 4-2 was used as input in assessing the existing shutdown strategy and consisted of the following activities: " Core methodology documents and plant specific calculations/analyses were gathered." Each specific section of NFPA 805 2.4.2 was correlated to the corresponding section of Chapter 3 of NEI 00-01 Revision 1. Based upon the content of the NEI 00-01 methodology statements, a determination was made of the applicability of the section to the station. The results of the applicability review were then documented in VCSNS Technical Report TR08620-014, "Nuclear Safety Performance Criteria Review Transition Report."" The plant-specific methodology was compared to applicable sections of NEI 00-01 and one of the following alignment statements and its associated basis were assigned to the section: o Aligns o Aligns with intent o Not in Alignment o Not in Alignment, but Prior NRC Approval o Not in Alignment, but no adverse consequences" For the existing shutdown analysis, no additional work was performed relative to corrective actions." Based on the VCSNS approach to transition the station shutdown analysis (FAQ 09-0057), project instructions and guidance were developed to analyze station circuits and documentation in accordance with NEI 00-01. A compliance review with appropriate references is contained in NEI 04-02 Table B-2 (Attachment B) to support the VCSNS transition to NFPA 805.Note: Since NEI 00-01 is a guidance document, portions of its text could be interpreted as 'good practice' or intended as an example of an efficient means of performing the analyses.The comparison of the VCSNS original safe shutdown methodology to NEI 00-01 Chapter 3 (NEI 04-02 Table B-2) was performed and documented in VCSNS Technical Compliance with NFPA 805 Requirements Page 17 Aýýo RC-11-0149 Section 4.0 Report TR08620-014.
(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" The NSCA methodology review evaluated the existing NSCA methodology against the guidance provided in NEI 00-01, Revision 1 Chapter 3, "Deterministic Methodology," as discussed in Appendix B-2 of NEI 04-02. The methodology in Figure 4-2 was used as input in assessing the existing shutdown strategy and consisted of the following activities: " Core methodology documents and plant specific calculations/analyses were gathered." Each specific section of NFPA 805 2.4.2 was correlated to the corresponding section of Chapter 3 of NEI 00-01 Revision 1. Based upon the content of the NEI 00-01 methodology statements, a determination was made of the applicability of the section to the station. The results of the applicability review were then documented in VCSNS Technical Report TR08620-014, "Nuclear Safety Performance Criteria Review Transition Report."" The plant-specific methodology was compared to applicable sections of NEI 00-01 and one of the following alignment statements and its associated basis were assigned to the section: o Aligns o Aligns with intent o Not in Alignment o Not in Alignment, but Prior NRC Approval o Not in Alignment, but no adverse consequences" For the existing shutdown analysis, no additional work was performed relative to corrective actions." Based on the VCSNS approach to transition the station shutdown analysis (FAQ 09-0057), project instructions and guidance were developed to analyze station circuits and documentation in accordance with NEI 00-01. A compliance review with appropriate references is contained in NEI 04-02 Table B-2 (Attachment B) to support the VCSNS transition to NFPA 805.Note: Since NEI 00-01 is a guidance document, portions of its text could be interpreted as 'good practice' or intended as an example of an efficient means of performing the analyses.The comparison of the VCSNS original safe shutdown methodology to NEI 00-01 Chapter 3 (NEI 04-02 Table B-2) was performed and documented in VCSNS Technical Compliance with NFPA 805 Requirements Page 17 Aýýo RC-11-0149 Section 4.0 Report TR08620-014.
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Note that the risk increase due to the use of recovery actions was included in the risk change for transition for each fire area.RG 1.205 Section C.2.2.4.2 states in part"The total increase or decrease in risk associated with the implementation of NFPA 805 for the overall plant should be calculated by summing the risk increases and decreases for each fire area (including any risk increases resulting from previously approved recovery actions).
Note that the risk increase due to the use of recovery actions was included in the risk change for transition for each fire area.RG 1.205 Section C.2.2.4.2 states in part"The total increase or decrease in risk associated with the implementation of NFPA 805 for the overall plant should be calculated by summing the risk increases and decreases for each fire area (including any risk increases resulting from previously approved recovery actions).
The total risk increase should be consistent with the acceptance guidelines in Regulatory Guide 1.174. Note that the acceptance guidelines of Regulatory Guide 1.174 may require the total CDF, LERF, or both, to evaluate changes where the risk impact exceeds specific guidelines.
The total risk increase should be consistent with the acceptance guidelines in Regulatory Guide 1.174. Note that the acceptance guidelines of Regulatory Guide 1.174 may require the total CDF, LERF, or both, to evaluate changes where the risk impact exceeds specific guidelines.
If the additional risk associated with previously approved recovery actions is greater than the acceptance guidelines in Regulatory Guide 1.174, then the net change in total plant risk incurred by any proposed alternatives to the deterministic criteria in NFPA 805, Chapter 4 (other than the previously approved recovery actions), should be risk neutral or represent a risk decrease." The risk increases and decreases are provided in Attachment W.4.6 Monitoring Program 4.6.1 Overview of NFPA 805 Requirements and NEI 04-02 Guidance on the NFPA 805 Fire Protection System and Feature Monitoring Program Section 2.6 of NFPA 805 states: "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." Compliance with NFPA 805 Requirements Page 49 Compliance with NFPA 805 Requirements Page 49 RC-11-0149 Section 4.0 The intent of the monitoring review is to confirm the adequacy of the surveillance, inspection, testing, compensatory measures, and oversight processes to support the Monitoring requirement defined in NFPA 805. This process considers the following: " Assessing the adequacy of the scope of structure, systems and components within existing plant programs" Defining the performance criteria for the availability and reliability of the required structure, systems and components" The adequacy of the plant corrective action program in determining causes of equipment and programmatic failures and in minimizing their recurrence 4.6.2 Overview of Post-Transition NFPA 805 Monitoring Program This section provides an overview of the post-transition NFPA 805 monitoring program process. The monitoring program will be implemented after the safety evaluation issuance as part of the fire protection program transition to NFPA 805. The monitoring program described in this section is currently based on FAQ 10-0059 Revision 1, which has not yet been issued a closure memo. VCSNS will implement a monitoring program consistent with the NRC approved version of FAQ 10-0059.The monitoring process is comprised of five phases." Phase 1 -Scoping" Phase 2 -Screening Using Risk Criteria" Phase 3 -Risk Target Value Determination" Phase 4 -Monitoring Implementation" Phase 5 -Periodic Assessment Phase 1 -Scoping In order to meet the NFPA 805 requirements for monitoring, the following categories of SSCs and programmatic elements will be reviewed during the implementation phase for inclusion in the NFPA 805 monitoring program: " Structures, Systems, and Components required to comply with NFPA 805, specifically:
If the additional risk associated with previously approved recovery actions is greater than the acceptance guidelines in Regulatory Guide 1.174, then the net change in total plant risk incurred by any proposed alternatives to the deterministic criteria in NFPA 805, Chapter 4 (other than the previously approved recovery actions), should be risk neutral or represent a risk decrease." The risk increases and decreases are provided in Attachment W.4.6 Monitoring Program 4.6.1 Overview of NFPA 805 Requirements and NEI 04-02 Guidance on the NFPA 805 Fire Protection System and Feature Monitoring Program Section 2.6 of NFPA 805 states: "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." Compliance with NFPA 805 Requirements Page 49 Compliance with NFPA 805 Requirements Page 49 RC-11-0149 Section 4.0 The intent of the monitoring review is to confirm the adequacy of the surveillance, inspection, testing, compensatory measures, and oversight processes to support the Monitoring requirement defined in NFPA 805. This process considers the following: " Assessing the adequacy of the scope of structure, systems and components within existing plant programs" Defining the performance criteria for the availability and reliability of the required structure, systems and components" The adequacy of the plant corrective action program in determining causes of equipment and programmatic failures and in minimizing their recurrence
 
====4.6.2 Overview====
of Post-Transition NFPA 805 Monitoring Program This section provides an overview of the post-transition NFPA 805 monitoring program process. The monitoring program will be implemented after the safety evaluation issuance as part of the fire protection program transition to NFPA 805. The monitoring program described in this section is currently based on FAQ 10-0059 Revision 1, which has not yet been issued a closure memo. VCSNS will implement a monitoring program consistent with the NRC approved version of FAQ 10-0059.The monitoring process is comprised of five phases." Phase 1 -Scoping" Phase 2 -Screening Using Risk Criteria" Phase 3 -Risk Target Value Determination" Phase 4 -Monitoring Implementation" Phase 5 -Periodic Assessment Phase 1 -Scoping In order to meet the NFPA 805 requirements for monitoring, the following categories of SSCs and programmatic elements will be reviewed during the implementation phase for inclusion in the NFPA 805 monitoring program: " Structures, Systems, and Components required to comply with NFPA 805, specifically:
o Fire protection systems and features required by the Nuclear Safety Capability Assessment o Fire protection systems and features modeled in the Fire PRA o Fire protection systems and features required by Chapter 3 of NFPA 805 o Nuclear Safety Capability Assessment equipment o Structures, systems and components relied upon to meet radioactive release criteria" Fire Protection Programmatic Elements" Key Assumptions in Engineering Analyses (specifically analyses performed to demonstrate compliance with the nuclear safety and radioactive release performance criteria)Compliance with NFPA 805 Requirements Page 50 Compliance with NFPA 805 Requirements Page 50 RC-11-0149 Section 4.0 As a minimum the fire protection systems and features and SSCs required to meet the radioactive release criteria will be included in the existing inspection and test programs and in the system/program health program. In addition passive features (rated barriers, ERFBS), and other components (e.g. drains, curbs) that are relied upon to demonstrate compliance with Chapter 4 of NFPA 805 will also be included in the inspection and test programs, including system/program health reporting.
o Fire protection systems and features required by the Nuclear Safety Capability Assessment o Fire protection systems and features modeled in the Fire PRA o Fire protection systems and features required by Chapter 3 of NFPA 805 o Nuclear Safety Capability Assessment equipment o Structures, systems and components relied upon to meet radioactive release criteria" Fire Protection Programmatic Elements" Key Assumptions in Engineering Analyses (specifically analyses performed to demonstrate compliance with the nuclear safety and radioactive release performance criteria)Compliance with NFPA 805 Requirements Page 50 Compliance with NFPA 805 Requirements Page 50 RC-11-0149 Section 4.0 As a minimum the fire protection systems and features and SSCs required to meet the radioactive release criteria will be included in the existing inspection and test programs and in the system/program health program. In addition passive features (rated barriers, ERFBS), and other components (e.g. drains, curbs) that are relied upon to demonstrate compliance with Chapter 4 of NFPA 805 will also be included in the inspection and test programs, including system/program health reporting.
The existing programs are adequate for routine monitoring of these SSCs. SSCs that are not addressed in the existing programs will be added.Phase 2 -Screening Using Risk Criteria Phase 2 of the process uses the risk significance criteria and screens the SSCs and programmatic elements to determine High Safety Significant SSCs and programmatic elements.
The existing programs are adequate for routine monitoring of these SSCs. SSCs that are not addressed in the existing programs will be added.Phase 2 -Screening Using Risk Criteria Phase 2 of the process uses the risk significance criteria and screens the SSCs and programmatic elements to determine High Safety Significant SSCs and programmatic elements.
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==5.0 REGULATORY EVALUATION==
==5.0 REGULATORY EVALUATION==


5.1 Introduction  
===5.1 Introduction===
 
-10 CFR 50.48 On July 16, 2004 the NRC amended 10 CFR 50.48, Fire Protection, to add a new subsection, 10 CFR 50.48(c), which establishes alternative fire protection requirements.
-10 CFR 50.48 On July 16, 2004 the NRC amended 10 CFR 50.48, Fire Protection, to add a new subsection, 10 CFR 50.48(c), which establishes alternative fire protection requirements.
10 CFR 50.48 endorses, with exceptions, NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants -2001 Edition, as a voluntary alternative for demonstrating compliance with 10 CFR 50.48 Section (b), Appendix R, and Section (f), Decommissioning.
10 CFR 50.48 endorses, with exceptions, NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants -2001 Edition, as a voluntary alternative for demonstrating compliance with 10 CFR 50.48 Section (b), Appendix R, and Section (f), Decommissioning.
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The italicized exception to Section 3.6.4 is See Section 4.1.2.2.not endorsed.
The italicized exception to Section 3.6.4 is See Section 4.1.2.2.not endorsed.
Licensees who wish to use the exception to Section 3.6.4 must submit a request for a license amendment in accordance with paragraph (c)(2)(vii) of this section.Regulatory Evaluation Page 87 Regulatory Evaluation Page 87 A-8 RC-11-0149 Section 5.0 Table 5-3 10 CFR 50.48(c) -Applicability/Compliance Reference 10 CFR 50.48(c) Section(s)
Licensees who wish to use the exception to Section 3.6.4 must submit a request for a license amendment in accordance with paragraph (c)(2)(vii) of this section.Regulatory Evaluation Page 87 Regulatory Evaluation Page 87 A-8 RC-11-0149 Section 5.0 Table 5-3 10 CFR 50.48(c) -Applicability/Compliance Reference 10 CFR 50.48(c) Section(s)
Applicability/Compliance Reference (vii) Performance-based methods. Notwithstanding the prohibition in Section The use of performance-3.1 against the use of performance-based methods, the fire protection based methods for NFPA 805 program elements and minimum design requirements of Chapter 3 may be Chapter 3 is requested.
Applicability/Compliance Reference (vii) Performance-based methods. Notwithstanding the prohibition in Section The use of performance-
 
===3.1 against===
the use of performance-based methods, the fire protection based methods for NFPA 805 program elements and minimum design requirements of Chapter 3 may be Chapter 3 is requested.
See subject to the performance-based methods permitted elsewhere in the Attachment L.standard.
See subject to the performance-based methods permitted elsewhere in the Attachment L.standard.
Licensees who wish to use performance-based methods for these fire protection program elements and minimum design requirements shall submit a request in the form of an application for license amendment under §50.90. The Director of the Office of Nuclear Reactor Regulation, or a designee of the Director, may approve the application if the Director or designee determines that the performance-based approach;(A) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release;(B) Maintains safety margins; and (C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).
Licensees who wish to use performance-based methods for these fire protection program elements and minimum design requirements shall submit a request in the form of an application for license amendment under §50.90. The Director of the Office of Nuclear Reactor Regulation, or a designee of the Director, may approve the application if the Director or designee determines that the performance-based approach;(A) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release;(B) Maintains safety margins; and (C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).
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Regulatory Evaluation Page 88 q -C"* RC-11-0149 Section 5.0 5.2 Regulatory Topics 5.2.1 License Condition Changes The current VCSNS fire protection license condition 2.c (18) is being replaced with the standard license condition based upon Regulatory Position 3.1 of RG 1.205, as shown in Attachment M.5.2.2 Technical Specifications VCSNS conducted a review of the Technical Specifications to determine which Technical Specifications are required to be revised, deleted, or superseded.
Regulatory Evaluation Page 88 q -C"* RC-11-0149 Section 5.0 5.2 Regulatory Topics 5.2.1 License Condition Changes The current VCSNS fire protection license condition 2.c (18) is being replaced with the standard license condition based upon Regulatory Position 3.1 of RG 1.205, as shown in Attachment M.5.2.2 Technical Specifications VCSNS conducted a review of the Technical Specifications to determine which Technical Specifications are required to be revised, deleted, or superseded.
VCSNS determined that the changes to the Technical Specifications and applicable justification listed in Attachment N are adequate for the VCSNS adoption of the new fire protection licensing basis.5.2.3 Orders and Exemptions A review was conducted of the VCSNS docketed correspondence to determine if there were any orders or exemptions that needed to be superseded or revised. A review was also performed to ensure that compliance with the physical protection requirements, security orders, and adherence to those commitments applicable to the plant are maintained.
VCSNS determined that the changes to the Technical Specifications and applicable justification listed in Attachment N are adequate for the VCSNS adoption of the new fire protection licensing basis.5.2.3 Orders and Exemptions A review was conducted of the VCSNS docketed correspondence to determine if there were any orders or exemptions that needed to be superseded or revised. A review was also performed to ensure that compliance with the physical protection requirements, security orders, and adherence to those commitments applicable to the plant are maintained.
A discussion of affected orders and exemptions is included in Attachment 0.5.3 Regulatory Evaluations 5.3.1 No Significant Hazards Consideration A written evaluation of the significant hazards consideration of a proposed license amendment is required by 10 CFR 50.92. According to 10 CFR 50.92, a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: " Involve a significant increase in the probability or consequences of an accident previously evaluated; or" Create the possibility of a new or different kind of accident from any accident previously evaluated; or" Involve a significant reduction in a margin of safety.This evaluation is contained in Attachment Q.Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. VCSNS has evaluated the proposed amendment and determined that it involves no significant hazards consideration.
A discussion of affected orders and exemptions is included in Attachment  
5.3.2 Environmental Consideration Pursuant to 10 CFR 51.22(b), an evaluation of the LAR has been performed to determine whether it meets the criteria for categorical exclusion set forth in 10 CFR Regulatory Evaluation Page 89  
 
====0.5.3 Regulatory====
 
Evaluations 5.3.1 No Significant Hazards Consideration A written evaluation of the significant hazards consideration of a proposed license amendment is required by 10 CFR 50.92. According to 10 CFR 50.92, a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: " Involve a significant increase in the probability or consequences of an accident previously evaluated; or" Create the possibility of a new or different kind of accident from any accident previously evaluated; or" Involve a significant reduction in a margin of safety.This evaluation is contained in Attachment Q.Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. VCSNS has evaluated the proposed amendment and determined that it involves no significant hazards consideration.
 
====5.3.2 Environmental====
 
Consideration Pursuant to 10 CFR 51.22(b), an evaluation of the LAR has been performed to determine whether it meets the criteria for categorical exclusion set forth in 10 CFR Regulatory Evaluation Page 89  
--a RC-11-0149 Section 5.0 51.22(c).
--a RC-11-0149 Section 5.0 51.22(c).
That evaluation is discussed in Attachment R. The evaluation confirms that this LAR meets the criteria set forth in 10 CFR 51.22(c)(9) for categorical exclusion from the need for an environmental impact assessment or statement.
That evaluation is discussed in Attachment R. The evaluation confirms that this LAR meets the criteria set forth in 10 CFR 51.22(c)(9) for categorical exclusion from the need for an environmental impact assessment or statement.
5.4 Transition Implementation Schedule The following schedule for transitioning VCSNS to the new fire protection licensing basis requires NRC approval of the LAR in accordance with the following schedule: " Implementation of new NFPA 805 fire protection program to include procedure changes, process updates, and training to affected plant personnel.
 
===5.4 Transition===
 
Implementation Schedule The following schedule for transitioning VCSNS to the new fire protection licensing basis requires NRC approval of the LAR in accordance with the following schedule: " Implementation of new NFPA 805 fire protection program to include procedure changes, process updates, and training to affected plant personnel.
This will occur one hundred eighty (180) days after NRC approval." Modifications scope and implementation schedule are provided in Attachment S. Appropriate compensatory measures will be maintained until modifications are complete.Regulatory Evaluation Page 90 Regulatory Evaluation Page 90 RC-11-0149 Section 6.0  
This will occur one hundred eighty (180) days after NRC approval." Modifications scope and implementation schedule are provided in Attachment S. Appropriate compensatory measures will be maintained until modifications are complete.Regulatory Evaluation Page 90 Regulatory Evaluation Page 90 RC-11-0149 Section 6.0  


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procedures to accomplish the following shall be established:
procedures to accomplish the following shall be established:
NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-4 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-4 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary (1) Inspection, testing, and maintenance CA SAP-0131, "Fire Protection Inspection, Testing and Maintenance for the for the protection systems and features Program", Rev 6D, Section Fire Protection Program are established in credited by the fire protection program 6.2.4, 6.2.5, 6.2.6 accordance with controlled procedures.
NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-4 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-4 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary (1) Inspection, testing, and maintenance CA SAP-0131, "Fire Protection Inspection, Testing and Maintenance for the for the protection systems and features Program", Rev 6D, Section Fire Protection Program are established in credited by the fire protection program 6.2.4, 6.2.5, 6.2.6 accordance with controlled procedures.
As FPP-023, "Fire Detection", Rev practical, performance-based surveillance 3A Enclosure 6.1 frequencies may be established as referenced FPP-024, "Fire Suppression", herein and described in Electric Power Rev 3C Enclosure 6.1 Research Institute (EPRI) technical report and FPP-025, "Fire Containment", NEIL Appendix 4.2.8. (Table S-2, Item 2)Rev 4F Enclosure 6.5, 6.6 EPRI Technical Report (TR)1006756 Fire Protection Surveillance Optimization and Maintenance Guide NEIL Appendix 4.2.8 (2) Compensatory actions implemented C FPP-023, "Fire Detection", Existing procedures address current when fire protection systems and other Rev 3A Enclosure 6.1 compensatory measures for the program, systems credited by the fire protection FPP-024, "Fire Suppression", however a new procedure establishing revised program and this standard cannot perform Rev 3C Enclosure 6.1 and updated compensatory measures will be their intended function and limits on FPP-025, "Fire Containment", developed during the implementation period to impairment duration Rev 4F Enclosure 6.5, 6.6 incorporate NFPA 805 insights. (Table S-2, Item 1)(3) Reviews of fire protection program- C SAP-0131, "Fire Protection Station Administrative Procedure (SAP)related performance and trends Program", Rev 6D establishes responsibilities for review of ES091 1, "Fire Protection program related performance and trends.Monitoring Program" Rev 0 System Engineering procedures establishes FP feature trends and monitoring. (Table S-2, Item 1) (Table S-2, Item 4)(4) Reviews of physical plant modifications C SAP-0133, "Design Control Engineering Services and Document Control and procedure changes for impact on the fire Program", Rev 14B Procedures manage interfaces and direct protection program SAP-0139, "Document Review appropriate documents to appropriate and Approval Process", Rev 32 personnel for FP program impacts. (Table S-2, ES-427 "Program / Issues Item 3)Screening" Rev 2D NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-5 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-5  
As FPP-023, "Fire Detection", Rev practical, performance-based surveillance 3A Enclosure  
 
===6.1 frequencies===
 
may be established as referenced FPP-024, "Fire Suppression", herein and described in Electric Power Rev 3C Enclosure  
 
===6.1 Research===
Institute (EPRI) technical report and FPP-025, "Fire Containment", NEIL Appendix 4.2.8. (Table S-2, Item 2)Rev 4F Enclosure 6.5, 6.6 EPRI Technical Report (TR)1006756 Fire Protection Surveillance Optimization and Maintenance Guide NEIL Appendix 4.2.8 (2) Compensatory actions implemented C FPP-023, "Fire Detection", Existing procedures address current when fire protection systems and other Rev 3A Enclosure  
 
===6.1 compensatory===
 
measures for the program, systems credited by the fire protection FPP-024, "Fire Suppression", however a new procedure establishing revised program and this standard cannot perform Rev 3C Enclosure 6.1 and updated compensatory measures will be their intended function and limits on FPP-025, "Fire Containment", developed during the implementation period to impairment duration Rev 4F Enclosure 6.5, 6.6 incorporate NFPA 805 insights. (Table S-2, Item 1)(3) Reviews of fire protection program- C SAP-0131, "Fire Protection Station Administrative Procedure (SAP)related performance and trends Program", Rev 6D establishes responsibilities for review of ES091 1, "Fire Protection program related performance and trends.Monitoring Program" Rev 0 System Engineering procedures establishes FP feature trends and monitoring. (Table S-2, Item 1) (Table S-2, Item 4)(4) Reviews of physical plant modifications C SAP-0133, "Design Control Engineering Services and Document Control and procedure changes for impact on the fire Program", Rev 14B Procedures manage interfaces and direct protection program SAP-0139, "Document Review appropriate documents to appropriate and Approval Process", Rev 32 personnel for FP program impacts. (Table S-2, ES-427 "Program / Issues Item 3)Screening" Rev 2D NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-5 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-5  
-ý RC-11-0149 Attachment A RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary (5) Long-term maintenance and C MD-21 "Configuration Long term maintenance and configuration of configuration of the fire protection program Management", Rev 7 the fire protection program is implemented SAP-1 31 "Fire Protection through a Management Directive (MD) and Program", Rev 6D Station Administrative Procedures (SAP).SAP-0139 "Document Review and Approval Process", Rev 32 SAP-0133, "Design Control Program", Rev 14B (6) Emergency response procedures for C EPP-013 "Fire Emergency", Station Emergency Response Procedures the plant industrial fire brigade Rev 8 direct plant actions for plant fire brigade response due to fires at the station.A fire prevention program with the goal of NRR FPP022 "Fire Prevention", Elements of the fire prevention program are preventing a fire from starting shall be Rev 3 described in the following subsections.
-ý RC-11-0149 Attachment A RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary (5) Long-term maintenance and C MD-21 "Configuration Long term maintenance and configuration of configuration of the fire protection program Management", Rev 7 the fire protection program is implemented SAP-1 31 "Fire Protection through a Management Directive (MD) and Program", Rev 6D Station Administrative Procedures (SAP).SAP-0139 "Document Review and Approval Process", Rev 32 SAP-0133, "Design Control Program", Rev 14B (6) Emergency response procedures for C EPP-013 "Fire Emergency", Station Emergency Response Procedures the plant industrial fire brigade Rev 8 direct plant actions for plant fire brigade response due to fires at the station.A fire prevention program with the goal of NRR FPP022 "Fire Prevention", Elements of the fire prevention program are preventing a fire from starting shall be Rev 3 described in the following subsections.
established, documented, and implemented as part of the fire protection program. The two basic components of the fire prevention program shall consist of both of the following:
established, documented, and implemented as part of the fire protection program. The two basic components of the fire prevention program shall consist of both of the following:
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The a listed fire-retardant application.
The a listed fire-retardant application.
Exception:
Exception:
TRP-02, "Fire Protection", Rev procurement of wood is incorporated as Fire Cribbing timbers 6 in. by 6 in. (15.2 cm by 15.2 8D Protection Program input into the procurement cm) or larger shall not be required to be fire- process.retardant treated.Untreated wood or lumber will be addressed for controls of limited duration, when required, as part of the fire protection program (e.g.outages, compensatory measures)(Table S-2, Item 1)NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-8 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-8 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary (2) Plastic sheeting materials used in the C FPP022 "Fire Prevention", Station procedures limit the type and use of power block shall be fire-retardant types that Rev 3 Enclosure 6.2 plastic sheeting materials at the station, to have passed NFPA 701, Standard Methods of TRP-02, "Fire Protection", Rev those with qualified test methods as described.
TRP-02, "Fire Protection", Rev procurement of wood is incorporated as Fire Cribbing timbers 6 in. by 6 in. (15.2 cm by 15.2 8D Protection Program input into the procurement cm) or larger shall not be required to be fire- process.retardant treated.Untreated wood or lumber will be addressed for controls of limited duration, when required, as part of the fire protection program (e.g.outages, compensatory measures)(Table S-2, Item 1)NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-8 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-8 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary (2) Plastic sheeting materials used in the C FPP022 "Fire Prevention", Station procedures limit the type and use of power block shall be fire-retardant types that Rev 3 Enclosure  
 
===6.2 plastic===
sheeting materials at the station, to have passed NFPA 701, Standard Methods of TRP-02, "Fire Protection", Rev those with qualified test methods as described.
Fire Tests for Flame Propagation of Textiles 8D Information concerning restrictions concerning and Films, large-scale tests, or equivalent, the procurement plastic sheeting has been addressed as Fire Protection inputs to the procurement process. (Table S-2, Item 1)(3) Waste, debris, scrap, packing materials, C SAP-0300 "Conduct of Station Maintenance procedures require or other combustibles shall be removed from Maintenance", Rev 12C, Section area cleanup following completion of an area immediately following the completion 6.1.17 maintenance activities. (Table S-2, Item 1)of work or at the end of the shift, whichever SAP-0142 "Station comes first. Housekeeping Program", Rev 15B (4)
Fire Tests for Flame Propagation of Textiles 8D Information concerning restrictions concerning and Films, large-scale tests, or equivalent, the procurement plastic sheeting has been addressed as Fire Protection inputs to the procurement process. (Table S-2, Item 1)(3) Waste, debris, scrap, packing materials, C SAP-0300 "Conduct of Station Maintenance procedures require or other combustibles shall be removed from Maintenance", Rev 12C, Section area cleanup following completion of an area immediately following the completion 6.1.17 maintenance activities. (Table S-2, Item 1)of work or at the end of the shift, whichever SAP-0142 "Station comes first. Housekeeping Program", Rev 15B (4)
* Combustible storage or staging areas C SAP-0142 "Station Station administrative procedures control the shall be designated, and limits shall be Housekeeping Program", Rev designation and management of Combustible established on the types and quantities of 15B Material Storage and Staging Areas (Table S-stored materials.
* Combustible storage or staging areas C SAP-0142 "Station Station administrative procedures control the shall be designated, and limits shall be Housekeeping Program", Rev designation and management of Combustible established on the types and quantities of 15B Material Storage and Staging Areas (Table S-stored materials.
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~iG., ~Attachment B NFPA 805 Section 2.4.2.1 Appendix B.1 Nuclear Safety Assessment The primary purpose of the nuclear safety assessment is to demonstrate that given cable and equipment damage due to a fire postulated in any fire area, sufficient equipment remains available to achieve the following nuclear safety performance criteria (see Section 1.5): (1) Reactivity control (2) Inventory and pressure control (3) Decay heat removal (4) Vital auxiliaries (5) Process monitoring The purpose of this appendix is to identify attributes that should be considered when demonstrating this capability.
~iG., ~Attachment B NFPA 805 Section 2.4.2.1 Appendix B.1 Nuclear Safety Assessment The primary purpose of the nuclear safety assessment is to demonstrate that given cable and equipment damage due to a fire postulated in any fire area, sufficient equipment remains available to achieve the following nuclear safety performance criteria (see Section 1.5): (1) Reactivity control (2) Inventory and pressure control (3) Decay heat removal (4) Vital auxiliaries (5) Process monitoring The purpose of this appendix is to identify attributes that should be considered when demonstrating this capability.
Other risk informed-performance-based methods acceptable to the AHJ are permitted.
Other risk informed-performance-based methods acceptable to the AHJ are permitted.
NEI 00-01 Section NEI 00-01 Section Description 3.1.2 Shutdown Functions 3.1.2.1 Reactivity Control 3.1.2.2 Pressure Control Systems The following discussion on each of these shutdown functions provides guidance for selecting the systems and equipment required for safe shutdown.
NEI 00-01 Section NEI 00-01 Section Description
 
====3.1.2 Shutdown====
Functions 3.1.2.1 Reactivity Control 3.1.2.2 Pressure Control Systems The following discussion on each of these shutdown functions provides guidance for selecting the systems and equipment required for safe shutdown.
For additional information on BWR system selection, refer to GE Report GENE- T43-00002-00 R01 entitled "Original Safe Shutdown Paths for the BWR."[BWR] Control Rod Drive System. The safe shutdown performance and design requirements for the reactivity control function can be met without automatic scram/trip capability.
For additional information on BWR system selection, refer to GE Report GENE- T43-00002-00 R01 entitled "Original Safe Shutdown Paths for the BWR."[BWR] Control Rod Drive System. The safe shutdown performance and design requirements for the reactivity control function can be met without automatic scram/trip capability.
Manual scram/reactor trip is credited.
Manual scram/reactor trip is credited.
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The set of systems and equipment to be considered for nuclear safety should address, as a minimum, the following:.(a) Systems and equipment required to place the plant in a safe and stable condition following a fire occurring while the plant is at power, or while maintaining hot standby or hot shutdown.
The set of systems and equipment to be considered for nuclear safety should address, as a minimum, the following:.(a) Systems and equipment required to place the plant in a safe and stable condition following a fire occurring while the plant is at power, or while maintaining hot standby or hot shutdown.
This fire also could result in a loss of off-site power, which would require achieving safe and stable conditions using power from on-site ac sources (i.e., emergency diesel generators).
This fire also could result in a loss of off-site power, which would require achieving safe and stable conditions using power from on-site ac sources (i.e., emergency diesel generators).
This is typically a traditional Appendix R to 10 CFR 50 post-fire safe shutdown analysis.(b) Systems and equipment required to maintain shutdown cooling capability following a fire originating while the plant is in the shutdown cooling mode.NEI 00-01 Section NEI 00-01 Section Description 3.1.3 Methodology for Shutdown System Selection 3.1.3.1 Identify Safe Shutdown Functions 3.1.3.2 Identify Combinations of Systems That Satisfy Each Safe Shutdown Function 3.1.3.3 Define Combination of Systems for Each Safe Shutdown Path 3.1.3.4 Assign Shutdown Paths to Each Combination of Systems Refer to Figure 3-2 for a flowchart illustrating the various steps involved in selecting safe shutdown systems and developing the shutdown paths. The following methodology may be used to define the safe shutdown systems and paths for an Appendix R analysis:  
This is typically a traditional Appendix R to 10 CFR 50 post-fire safe shutdown analysis.(b) Systems and equipment required to maintain shutdown cooling capability following a fire originating while the plant is in the shutdown cooling mode.NEI 00-01 Section NEI 00-01 Section Description
 
====3.1.3 Methodology====
 
for Shutdown System Selection 3.1.3.1 Identify Safe Shutdown Functions 3.1.3.2 Identify Combinations of Systems That Satisfy Each Safe Shutdown Function 3.1.3.3 Define Combination of Systems for Each Safe Shutdown Path 3.1.3.4 Assign Shutdown Paths to Each Combination of Systems Refer to Figure 3-2 for a flowchart illustrating the various steps involved in selecting safe shutdown systems and developing the shutdown paths. The following methodology may be used to define the safe shutdown systems and paths for an Appendix R analysis:  
[Refer to hard copy of NEI 00-01 for Figure 3-2]Review available documentation to obtain an understanding of the available plant systems and the functions required to achieve and maintain safe shutdown.Documents such as the following may be reviewed:
[Refer to hard copy of NEI 00-01 for Figure 3-2]Review available documentation to obtain an understanding of the available plant systems and the functions required to achieve and maintain safe shutdown.Documents such as the following may be reviewed:
* Operating Procedures (Normal, Emergency, Abnormal);
* Operating Procedures (Normal, Emergency, Abnormal);
Line 1,106: Line 1,162:
-Methodology Review Page B-26 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment  
-Methodology Review Page B-26 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment  
-Methodology Review Page B-26 Attachment B NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis 3.3.1.7 [Circuit Coordination]
-Methodology Review Page B-26 Attachment B NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis 3.3.1.7 [Circuit Coordination]
3.3.2 Associated Circuit Cables 3.3.2 [A, Cables Whose Failure May Cause Spurious Actuations]
 
====3.3.2 Associated====
 
Circuit Cables 3.3.2 [A, Cables Whose Failure May Cause Spurious Actuations]
3.3.2 [B, Common Power Source Cables]3.3.2. [C, Common Enclosure Cables]Cabling for the electrical distribution system is a concern for those breakers that feed associated circuits and are not fully coordinated with upstream breakers.
3.3.2 [B, Common Power Source Cables]3.3.2. [C, Common Enclosure Cables]Cabling for the electrical distribution system is a concern for those breakers that feed associated circuits and are not fully coordinated with upstream breakers.
With respect to electrical distribution cabling, two types of cable associations exist. For safe shutdown considerations, the direct power feed to a primary safe shutdown component is associated with the primary component.
With respect to electrical distribution cabling, two types of cable associations exist. For safe shutdown considerations, the direct power feed to a primary safe shutdown component is associated with the primary component.
Line 1,168: Line 1,227:
The consideration of more than two cables being damaged (and subsequent spurious actuations) is deferred pending additional research.For cases involving direct current (DC) circuits, the potential spurious operation due to failures of the associated control cables (even if the spurious operation requires two concurrent hot shorts of the proper polarity, e.g., plus-to-plus and minus-to-minus) should be considered when the required source and target conductors are each located within the same multiconductor cable.Aligns Aligns See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 Aligns Aligns See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 Aligns NEI 04-02 Table B-2 Nuclear Safety Capability Assessment  
The consideration of more than two cables being damaged (and subsequent spurious actuations) is deferred pending additional research.For cases involving direct current (DC) circuits, the potential spurious operation due to failures of the associated control cables (even if the spurious operation requires two concurrent hot shorts of the proper polarity, e.g., plus-to-plus and minus-to-minus) should be considered when the required source and target conductors are each located within the same multiconductor cable.Aligns Aligns See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 Aligns Aligns See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 Aligns NEI 04-02 Table B-2 Nuclear Safety Capability Assessment  
-Methodology Review Page B-30 "lewsm Attachment B NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis 3.5.1.5 [E, Instrumentation Circuits]3.5.1.5 [F, Undesired Consequences]
-Methodology Review Page B-30 "lewsm Attachment B NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis 3.5.1.5 [E, Instrumentation Circuits]3.5.1.5 [F, Undesired Consequences]
3.5.2 Types of Circuit Failures Instrumentation Circuits.
 
====3.5.2 Types====
of Circuit Failures Instrumentation Circuits.
Required instrumentation circuits are beyond the scope of this associated circuit approach and must meet the same requirements as required power and control circuits.
Required instrumentation circuits are beyond the scope of this associated circuit approach and must meet the same requirements as required power and control circuits.
There is one case where an instrument circuit could potentially be considered an associated circuit. If fire-induced damage of an instrument circuit could prevent operation (e.g., lockout permissive signal) or cause maloperation (e.g., unwanted start/stop/reposition signal) of systems necessary to achieve and maintain hot shutdown, then the instrument circuit may be considered an associated circuit and handled accordingly.
There is one case where an instrument circuit could potentially be considered an associated circuit. If fire-induced damage of an instrument circuit could prevent operation (e.g., lockout permissive signal) or cause maloperation (e.g., unwanted start/stop/reposition signal) of systems necessary to achieve and maintain hot shutdown, then the instrument circuit may be considered an associated circuit and handled accordingly.

Revision as of 16:02, 13 October 2018

Virgil C. Summer, Unit 1 - License Amendment Request - LAR-06-00055 - License Amendment Request to Adopt NFPA 805 Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition). Part 1 of 2
ML11332A077
Person / Time
Site: Summer South Carolina Electric & Gas Company icon.png
Issue date: 11/15/2011
From: Gatlin T D
SCANA Corp, South Carolina Electric & Gas Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR-06-00055 RC-11-0149
Download: ML11332A077 (443)
v • d • e


Text

lnformztion Withhold frotam Publia Dicolc9c Undcr 10 GFR 2.350 Thomas D. Gatlin Vice President, Nuclear Operations 803.345.4342 A SCANA COMPANY November 15, 2011 RC-1 1-0149 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Dear Sir / Madam:

Subject:

VIRGIL C. SUMMER NUCLEAR STATION UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 LICENSE AMENDMENT REQUEST -LAR-06-00055 LICENSE AMENDMENT REQUEST TO ADOPT NFPA 805 PERFORMANCE-BASED STANDARD FOR FIRE PROTECTION FOR LIGHT WATER REACTOR ELECTRIC GENERATING PLANTS (2001 EDITION)Pursuant to 10 CFR 50.90, South Carolina Electric & Gas Company (SCE&G), acting for itself and as agent for South Carolina Public Service Authority, requests an amendment to the Virgil C. Summer Nuclear Station (VCSNS) Unit 1 Facility Operating License No. NFP-12. This LAR requests the Nuclear Regulatory Commission (NRC) review and approval for adoption of a new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a), 10 CFR 50.48(c), and the guidance in Regulatory Guide (RG) 1.205, Revision 1. The LAR also follows the applicable guidance in Nuclear Energy Institute (NEI) 04-02, Revision 2.The transition includes the following high level activities:

elimination of the Self-Induced Station Blackout (SISBO) methodology; a new Nuclear Safety Capability Assessment (NSCA) to replace the Appendix R safe shutdown analysis; a new Fire Probabilistic Risk Assessment (Fire PRA); a new Non-Power Operations (NPO) assessment; a new Radiological Release assessment; and completion of activities required for transitioning the licensing basis to 10 CFR 50.48(c) as specified in NEI 04-02 and RG 1.205.The NFPA 805 Task Force, established by NEI to ensure successful implementation of NFPA 805 consistent with RG 1.205, continues to provide the interface between the transitioning plants, the nuclear industry, and the NRC. The Task Force, working with the NRC, developed and maintains the Frequently Asked Questions (FAQ) process for obtaining clarifications to RG 1.205, NEI 04-02, and NFPA 805. Additional information is provided in Section 3.4 of Enclosure 1. Attachment H of Enclosure 1 provides the FAQs to date that have been reviewed and/or used to clarify the guidance listed above. FAQ is an ongoing process that will continue as licensees transition.

Attachments C, D, G, and W of Enclosure I transmitted herewith contain sensitive information.

When separated from these enclosures, this transmittal document is decontrolled.

KT _11I;LýýIIIL -a r.a4i w_=Tm~~ii q'm i- I ~ ~ ~ sr Je t_-_ .. --M-14-OMV

i ;;z_
;;Virgil C. Summer Station .Post Office Box 88
  • Jenkinsville, SC .29065
  • T (803) 345-5209 A~~ILV IE I I r- 14L@JI ýT V I qI IIIJL-1 V M L-AI IN W IY -J I.U6 LM -A Idn vli-I..3JUT Z- iZ7 Document Control Desk CR-06-00055 RC-11-0149 Page 2 of 3 Enclosure 1 contains the VCSNS Transition Report and its supporting attachments.

The Transition Report provides the required technical and regulatory assessments to enable the NRC to complete the review and approval of the new licensing basis. SCE&G considers Attachments C, D, G, and W of the Transition Report to be sensitive information and requests that it be withheld from public disclosure pursuant to 10 CFR 2.390.The Fire PRA to support the Risk-Informed, Performance-Based (RI-PB) fire risk evaluations per Regulatory Positions 2.2 and 4.3 of RG 1.205 has been completed.

The Fire PRA was developed in accordance with ASME/ANS RA-Sa-2009 and the guidance in NUREG/CR-6850/EPRI TR-1011989 and the NFPA FAQs. A peer review was conducted during the period of August 16, 2010 through August 20, 2010, and a follow-on peer review was conducted the week of February 21, 2011. This is further discussed in Section 4.5 of Enclosure 1.A number of variances were identified during the development of the NFPA 805 Nuclear Safety Capability Assessment and dispositioned using performance-based methods. These methods include fire modeling (NFPA 805, Section 4.2.4.1) and fire risk evaluation (NFPA 805, Section 4.2.4.2) processes.

Variances were assessed against the quantitative risk acceptance criteria and maintenance of defense-in-depth, and safety margin criteria were ensured as required by Section 5.3.5 of NEI 04-02 and RG 1.205. The results are summarized in Attachment C of Enclosure 1.Operator manual actions (OMAs) will be described as "recovery actions" in the new licensing basis. As a result of the elimination of SISBO compliance strategy, only a limited number of pre-transition OMAs were retained and no new recovery actions were added. The remaining recovery and primary control station actions are associated with Control Complex fires, when Control Room evacuation is required.

Section 4.2.1.3 and Attachment G of Enclosure 1 discuss the methodology and results associated with treatment of OMAs.Attachment S of Enclosure 1 contains a list of plant modifications and implementation items to support transitioning to the new fire protection licensing basis.Enclosure 2 contains the VCSNS List of Regulatory Commitments related to the transition to NFPA 805.Enclosure 3 contains the VCSNS Operating License and Technical Specification changes related to the transition to NFPA 805.SCE&G has notified the State of South Carolina in accordance with 10 CFR 50.91.Upon approval, SCE&G requests implementation of the amendment to occur within 180 days of approval.If you have any questions or require additional information, please contact Bruce Thompson at (803) 931-5042.

t i.anrm at;h S 1-armenIeIniRFI

  • IIi II... i. vis..3Lt u 1U it 2.asa Document Control Desk CR-06-00055 RC-1 1-0149 Page 3 of 3 I certify under penalty of perjury that the information contained herein is true and correct.11- 14e _Z0 Executed on Thomas D. Gatlin GAR/TDGIjg Attachment(s):

Enclosure 1 -Transition Report Enclosure 2 -List of Commitments Enclosure 3 -Operating License & Technical Specification Changes c: Without Attachments Unless Noted K. B. Marsh S. A. Byrne J. B. Archie N. S. Cams J. H. Hamilton R. J. White W. M. Cherry V. M. McCree W B. C. Gleaves W NRC Resident Inspector K. M. Sutton S. E. Jenkins W P. Ledbetter NSRC RTS (CR-06-00055)

File (813.20)PRSF (RC-11-0149) ith Attachments ith Attachments ith Attachments fith Attachments RC-1 1-0149 Enclosure 1 South Carolina Electric & Gas Company Virgil C. Summer Nuclear Station Docket 50-395 Transition to 10 CFR 50.48(c) -NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition A SCANA COMPAJE Transition Report November 15, 2011 RC-11-0149 TABLE OF CONTENTS Executive Summary .................................................................................................

iv A cronym List ................................................................................................................

vi

1.0 INTRODUCTION

.....................................................................................................

1 1 .1 B a ckg ro u n d ........................................................................................................

1 1.1.1 NFPA 805 -Requirements and Guidance .................................................

1 1.1.2 Transition to 10 CFR 50.48(c) ...............................................................

2 1.2 P urpose ...............................................................................................

.....3 2.0 OVERVIEW OF EXISTING FIRE PROTECTION PROGRAM ............................

4 2.1 Current Fire Protection Licensing Basis ........................................................

4 2.2 NRC Acceptance of the Fire Protection Licensing Basis ...............................

4 3.0 TRANSITION PROCESS .....................................................................................

7 3 .1 B ackg ro u nd ..................................................................................................

..7 3.2 N FPA 805 Process ........................................................................................

7 3.3 NEI 04-02 -NFPA 805 Transition Process ....................................................

8 3.4 NFPA 805 Frequently Asked Questions (FAQs) ...........................................

9 4.0 COMPLIANCE WITH NFPA 805 REQUIREMENTS

........................................

11 4.1 Fundamental Fire Protection Program and Design Elements ......................

11 4.1.1 Overview of Evaluation Process ..........................................................

11 4.1.2 Results of the Evaluation Process ......................................................

13 4.1.3 Definition of Power Block and Plant ....................................................

16 4.2 Nuclear Safety Performance Criteria ..........................................................

16 4.2.1 Nuclear Safety Capability Assessment Methodology

...........................

16 4.2.2 Fire Protection Engineering Equivalency Evaluation (FPEEE) Transition 29 4.2.3 Licensing Actions: Resulting NFPA 805 Analysis & Transitions

....... 30 4.2.4 Fire A rea D isposition

............................................................................

33 4.3 Non-Power Operational Modes ...................................................................

37 4.3.1 Overview of Evaluation Process ..........................................................

37 4.3.2 Results of the Evaluation Process ......................................................

39 4.4 Radioactive Release Performance Criteria .................................................

40 4.4.1 Overview of Evaluation Process ..........................................................

40 4.4.2 Results of the Evaluation Process ......................................................

41 4.5 Fire PRA and Performance-Based Approaches

..........................................

41 4.5.1 Fire PRA Development and Assessment

.............................................

42 4.5.2 Performance-Based Approaches

........................................................

44 NFPA 805 Transition Report Pagei1

-V RC-11-0149 4.6 M onitoring Program ....................................................................................

49 4.6.1 Overview of NFPA 805 Requirements and NEI 04-02 Guidance on the NFPA 805 Fire Protection System and Feature Monitoring Program ...... 49 4.6.2 Overview of Post-Transition NFPA 805 Monitoring Program ...............

50 4.7 Program Documentation, Configuration Control, and Quality Assurance

........ 53 4.7.1 NFPA 805 Documentation Requirements (NFPA 805, Section 2.7.1) ..... 53 4.7.2 NFPA 805 Configuration Management (NFPA 805, Sections 2.2.9/2.7.2 55 4.7.3 NFPA 805 Quality Requirements (NFPA 805, Section 2.7.3) ..............

58 4.8 Sum m ary of R esults .....................................................................................

60 4.8.1 Results of the Fire Area Review ..........................................................

60 4.8.2 Required Fire Protection System/Feature

...........................................

63 4.8.3 Fire R isk Insights ................................................................................

80 4.8.4 Plant Modifications and Items to be Completed During the Implementation P ha se ................................................................................................

..8 1 4.9 Supplemental Information

-Other VCSNS Specific Issues ........................

82 4.9.1 Self-Induced Station Blackout (SISBO) ................................................

82 4.9.2 NFPA 805 Chapter 4 Requirements for Approval ................................

83

5.0 REGULATORY EVALUATION

........................................................................

84 5.1 Introduction-10 CFR 50.48 .......................................................................

84 5.2 R egulatory Topics .......................................................................................

89 5.2.1 License Condition Changes .................................................................

89 5.2.2 Technical Specifications

.....................................................................

89 5.2.3 Orders and Exemptions

.......................................................................

89 5.3 Regulatory Evaluations

................................................................................

89 5.3.1 No Significant Hazards Consideration

.................................................

89 5.3.2 Environmental Consideration

...............................................................

89 5.4 Transition Implementation Schedule ..........................................................

90

6.0 REFERENCES

..................................................................................................

91 ATTACHMENTS

......................................................................................................

94 A. NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program &Design Elements .........................................................................................

A-1 B. NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology R eview ........................................................................................................

B -1 C. NEI 04-02 Table B-3 Fire Area Transition

.................................................

C-1 D. NEI 04-02 Non-Power Operational Modes Transition

..............................

D-1 E. NEI 04-02 Radioactive Release Transition

................................................

E-1 NFPA 805 Transition Report Page ii NFPA 805 Transition Report Page ii A-8 RC-11-0149 F. Fire-induced Multiple Spurious Operations Resolution

...........................

F-I G. Recovery Actions Transition

.....................................................................

G-I H. NFPA 805 Frequently Asked Question Summary Table ..........................

H-1 I. Definition of Power Block ............................................................................

1-I J. Fire M odeling V&V .......................................................................................

J-1 K. Existing Licensing Action Transition

.......................................................

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

L-I M. License Condition Changes .....................................................................

M-1 N. Technical Specification Changes .............................................................

N-I 0. Orders and Exemptions

............................................................................

0-I P. RI-PB Alternatives to NFPA 805 10 CFR 50.48(c)(4)

................................

P-I Q. No Significant Hazards Evaluations

.......................................................

Q-I R. Environmental Considerations Evaluation

................................................

R-I S. Plant Modifications and Items to be Completed During Implementation

...S-I T. Clarification of Prior NRC Approvals

........................................................

T-1 U. Internal Events PRA Quality .....................................................................

U-1 V. Fire PRA Q uality .........................................................................................

V-I W. Fire PRA Insights .......................................................................................

W-I X. Other Requests for Approval .....................................................................

X-I NFPA 805 Transition Report Page iii NFPA 805 Transition Report Page iii RC-11-0149 Executive Summary South Carolina Electric & Gas Company (SCE&G) will transition the Virgil C. Summer Nuclear Station (VCSNS) fire protection program and the current licensing basis (CLB)to a new Risk Informed-Performance Based (RI-PB) alternative per 10 CFR 50.48(c), which incorporates by reference NFPA 805. The CLB per 10 CFR 50.48(b) and 10 CFR 50, Appendix R, which has been in place since the early 1980s will be superseded in its entirety.In 2006, SCE&G elected to adopt NFPA 805. A letter of intent was submitted by SCE&G to the NRC on October 19, 2006 (ML062990453) for VCSNS to adopt NFPA 805 in accordance with 10 CFR 50.48(c).

The letter of intent requested three years of enforcement discretion.

The NRC responded to SCE&G on January 19, 2007 (ML063520409) and approved an enforcement discretion period from October 19, 2006 to October 19, 2009. On July 16, 2009, SCE&G submitted a request in accordance with COMSECY-08-022 to extend the enforcement discretion to six months past the date of the safety evaluation approving the second pilot plant LAR review. In a letter (ML092920297) dated October 19, 2009, the NRC granted this enforcement discretion extension request. However, due to the large number of LAR submittals expected in June 2011, in a letter (ML1 116106160) dated June 10, 2011, the Commission approved the staffs recommendation to publish the Federal Register Notice (FRN) announcing the revision to the Enforcement Policy to extend the enforcement discretion to correspond with a staggered LAR submittal schedule.

On June 23, 2011, SCE&G submitted a letter (RC-1 1-0099) requesting extension of their enforcement discretion and committed to the submittal date of September 30, 2011. In a letter (RC-11-0161) dated September 30, 2011, SCE&G informed the NRC that additional evaluation and clarification was needed to ensure the Transition Report met the completeness expectation, and that the LAR will be submitted by November 17, 2011.The transition process consisted of a review and update of VCSNS documentation, including the development of a Fire Probabilistic Risk Assessment (Fire PRA) using NUREG/CR 6850 as guidance.

This Transition Report summarizes the transition process and results. This Transition Report contains information: " Required by 10 CFR 50.48(c)." Recommended by guidance document Nuclear Energy Institute (NEI) 04-02 Revision 2 and appropriate Frequently Asked Questions (FAQs)." Recommended by guidance document Regulatory Guide 1.205 Revision 1.Section 4 of the Transition Report provides a summary of compliance with the following NFPA 805 requirements: " Fundamental Fire Protection Program Elements and Minimum Design Requirements" Nuclear Safety Performance Criteria, including:

o Non-Power Operational Modes o Fire Risk Evaluations" Radioactive Release Performance Criteria Executive Summary Page iv RC-11-0149" Monitoring Program" Program Documentation, Configuration Control, and Quality Assurance Section 5 of the Transition Report provides regulatory evaluations and associated attachments, including: " Changes to License Condition" Changes to Technical Specifications, Orders, and Exemptions," Determination of No Significant Hazards and evaluation of Environmental Considerations.

The attachments to the Transition Report include detail to support the transition process and results.Attachment H contains the approved FAQs not yet incorporated into the endorsed revision of NEI 04-02. These FAQs have been reviewed and/or used to clarify the guidance in RG 1.205, NEI 04-02, and the requirements of NFPA 805 and in the preparation of this License Amendment Request.Executive Summary Page v Executive Summary Page v A C e RC-11-0149 Acronym List ACDF -Change in CDF ALERF -Change in LERF AC -Alternating Current ADAMS -Agencywide Documents Access and Management System AF -Auxiliary Feedwater AHJ -Authority Having Jurisdiction ALARA -As Low As Reasonably Achievable ANS -American Nuclear Society APCSB -Auxiliary Power Conversion Systems Branch ARC -VCSNS Fire Safe Shutdown Compliance Assessment Program ASI -Alternate Seal Injection ASME -American Society of Mechanical Engineers ATWS -Anticipated Transient Without Scram BTP -Branch Technical Position CAFTA -Computer Aided Fault Tree Analysis CBDTM -Cause Based Decision Tree Methodology CC -Capability Category / Component Cooling CCDP -Conditional Core Damage Probability CCF -Common Cause Failure CCFA -Common Cause Failure Analysis CCW -Condenser Cooling Water CDF -Core Damage Frequency CF -Circuit Failure CFAST -Consolidated Fire and Smoke Transport Model CFR -Code of Federal Regulation CLB -Current Licensing Basis CREP -Control Room Evacuation Panel CS -Cable Selection CST -Condensate Storage Tank CVCS -Chemical and Volume Control System DBA -Design Basis Accident Acronym List Page vi Acronym List Page vi AE -1 RC-11-0149 DBD -Design Basis Document DC -Direct Current DH -Decay Heat DID -Defense-in-Depth DROID -Deterministic Requirement Open Item Description ECCS -Emergency Core Cooling System ECR -Engineering Change Request EDG -Emergency Diesel Generator EFW -Emergency Feedwater System EOP -Emergency Operating Procedure EP -Radiation Emergency Plan EPP -Emergency Plan Procedure EPRI -Electric Power Research Institute ERFBS -Electrical Raceway Fire Barrier System ES -Equipment Selection ESFAS -Engineered Safeguards Features Actuation Signals F&O -Fact and Observation FAQ -Frequently Asked Question FDS -Fire Dynamics Simulator FDT -Fire Dynamic Tool FEP -Fire Emergency Procedure Fire PRA -Fire Probabilistic Risk Assessment FM -Fire Modeling / Factory Mutual FP -Fire Protection FPEEE -Fire Protection Engineering Equivalency Evaluation FPER -Fire Protection Evaluation Report FPP -Fire Protection Procedure FQ -Fire Risk Quantification FRE -Fire Risk Evaluation FRN -Federal Register Notice FSAR -Final Safety Analysis Report FSS -Fire Scenario Selection GDC -General Design Criterion Acronym List Page vii "M RC-11-0149 HEP -Human Error Probability HGL -Hot Gas Layer HLP -Hi-Low Pressure HLR -High Level Requirement HRA -Human Reliability Analysis HRE -Higher Risk Evolution HSS -High Safety Significant HVAC -Heating, Ventilation, and Air Conditioning IC -Inventory Control IEEE -Institute of Electrical and Electronics Engineers IGN -Ignition Frequency INEEL -Idaho National Engineering and Environmental Laboratory IPE -Individual Plant Examination ISLOCA -Interfacing-Systems Loss of Coolant Accident KSF -Key Safety Function LAR -License Amendment Request LCC -Total Loss of CCW System LERF -Large Early Release Frequency LFS -Limiting Fire Scenario LOCA -Loss of Coolant Accident LP -Low Pressure LPI -Low Pressure Injection MCC -Motor Control Center MCR -Main Control Room MD -Management Directive MEFS -Maximum Expected Fire Scenario MSO -Multiple Spurious Operation MU -Maintenance and Update NEI 04-02 -NEI 04-02, "Guidance for Implementing a Risk-informed, Performance-based Fire Protection Program Under 10 CFR 50.48(c)" NEIL -Nuclear Energy Insurance Limited NFPA 805 -National Fire Protection Association Standard 805 NHT -National Hose Thread Acronym List Page viii RC-11-0149 NRC -Nuclear Regulatory Commission NPO -Non-Power Operations NSA -Nuclear Safety Assessment NSCA -Nuclear Safety Capability Assessment NSEL -Nuclear Safety Equipment List NSP -Nuclear Safety Performance OAP -Operations Administrative Procedure OMA -Operatory Manual Action OS&Y -Outside Stem & Yoke OSHA -Occupational Safety and Health Administration PC-CKS -Electrical Cable and Raceway Management System PCS -Primary Control Station PORV -Power Operated Relief Valve POS -Plant Operational State PP -Plant Partitioning PRA -Probabilistic Risk Assessment PRM -Plant Response Model Psacd -Probability of Spurious Actuation given Cable Damage PSV -Pressurizer Safety Valve PTP -Preventative Test Procedure PWROG -Pressurized Water Reactor Owners Group QNS -Quantitative Screening QSP -Quality Systems Procedure RAW -Risk Achievement Worth RA -Recovery Action RES -NRC Office of Nuclear Regulatory Research RC -Reactivity Control RCA -Radioactive Control Area RCP -Reactor Coolant Pump RCS -Reactor Coolant System RHR -Residual Heat Removal RIS -Regulatory Issues Summary RG -Regulatory Guide Acronym List Page ix RC-11-0149 RI-PB -Risk-Informed, Performance-Based RIS -Regulatory Issues Summary RWST -Refueling Water Storage Tank SAP -Station Administrative Procedure SBO -Station Blackout SCE&G -South Carolina Electric & Gas Company SER -Safety Evaluation Report SF -Seismic Fire SG -Steam Generator SISBO -Self-Induced Station Blackout SOE -Spurious Operation of Equipment SP -Specification SR -Standard Support Requirements (PRA standard reference)

SRP -Standard Review Plan SSCA -Safe Shutdown Circuit Analysis SSCs -Structures, Systems, and Components SSD -Appendix R Safe Shutdown SSE -Safe Shutdown Earthquake SSER -Supplemental Safety Evaluation Report STP -Surveillance Test Procedure SW -Service Water TAC -Technical Assignment Control TD -Turbine-Driven TH -Thermal Hydraulic TQP -Training and-Qualification Procedure UFSAR -Updated Final Safety Assessment Report UL -Underwriters Laboratories UNC -Uncertainty and Sensitivity V&V -Verification

& Validation VA -Vital Auxiliaries VCSNS -Virgil C. Summer Nuclear Station VFDR -Variance from Deterministic Requirement WOG -Westinghouse Owners Group Acronym List Page x Acronym List Page x

-ý° RC-11-0149 Section 1.0

1.0 INTRODUCTION

The Nuclear Regulatory Commission (NRC) has promulgated an alternative rule for fire protection requirements at nuclear power plants, 10 CFR 50.48(c), National Fire Protection Association Standard 805 (NFPA 805), 2001 Edition. South Carolina Electric& Gas Company (SCE&G) is implementing the Nuclear Energy Institute methodology NEI 04-02, Revision 2, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c)" (NEI 04-02), to transition Virgil C. Summer Nuclear Station (VCSNS) from its current fire protection licensing basis to the new requirements as outlined in NFPA 805. This report describes the transition methodology utilized and documents how VCSNS complies with the new requirements.

1.1 Background

1.1.1 NFPA 805- Requirements and Guidance On July 16, 2004 the NRC amended 10 CFR 50.48, Fire Protection, to add a new subsection, 10 CFR 50.48(c), which establishes new Risk-Informed, Performance-Based (RI-PB) fire protection requirements.

10 CFR 50.48(c) incorporates by reference, with exceptions, the National Fire Protection Association's NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants -2001 Edition, as a voluntary alternative to 10 CFR 50.48 Section (b), Appendix R, and Section (f), Decommissioning.

As stated in 10 CFR 50.48(c)(3)(i), any licensee's adoption of a RI-PB program that complies with the rule is voluntary.

This rule may be adopted as an acceptable alternative method for complying with either 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, or 10 CFR 50.48(f), plants shutdown in accordance with 10 CFR 50.82(a)(1).

NEI developed NEI 04-02 to assist licensees in adopting NFPA 805 and making the transition from their current fire protection licensing basis to one based on NFPA 805.The NRC issued Regulatory Guide (RG) 1.205, Risk-Informed, Performance-Based Fire Protection for Existing Light Water Nuclear Power Plants, which endorses NEI 04-02, with exceptions, in December 2009.1 A depiction of the primary document relationships is shown in Figure 1-1: 1 Where referred to in this document NEI 04-02 is Revision 2 and RG 1.205 is Revision 1.Introduction Page 1 Introduction Page 1 Aý8 RC-11-0149 Section 1.0 Incorporation by E a-0)Perfermanee Based StAndar~dfor FIRfor PI LqtWa pa-n-Incorporation by Reference z Z 10 CFR 50.48(c)National Fire Protection Association

_- Standard NFPA 805 EEEFMý'CD NEI 04-02 GUIDANCE FOR IMPLEMENTING A RI-PB FP PROGRAM UNDER 10 CFR 50.48(c)Endorsement RG 1.205 RI-PB FP FOR EXISTING LIGHT-WATER NUCLEAR POWER PLANTS Figure 1-1 NFPA 805 Transition

-Implementation Requirements/Guidance

1.1.2 Transition

to 10 CFR 50.48(c)1.1.2.1 Start of Transition In October 2006, VCSNS decided to transition the fire protection licensing basis to the RI-PB alternative in 10 CFR 50.48(c).

SCE&G submitted a letter of intent to the NRC on October 19, 2006 (ML062990453) for VCSNS to adopt NFPA 805 in accordance with 10 CFR 50.48(c).By letter dated January 19, 2007 (ML063520409), the NRC granted a three year enforcement discretion period from October 19, 2006 to October 19, 2009. In accordance with NRC Enforcement Policy, the enforcement discretion period will continue until the NRC approval of the license amendment request (LAR) is completed.

On July 16, 2009 SCE&G submitted a request in accordance with COMSECY-08-022 to extend the enforcement discretion to six months past the date of the safety evaluation approving the second pilot plant LAR review. In a letter (ML092920297) dated October 19, 2009, the NRC granted this enforcement discretion extension request.The NRC expected approximately 23 LARs by the end of June 2011. As a result, the Commission worked with industry to develop and create a staggered LAR submittal schedule.

On April 14, 2011, the NRC held a public meeting, during which the staff and Introduction Page 2 RC-11-0149 Section 1.0 stakeholders discussed the staggered approach method. In a letter (ML1 11101452)dated April 20, 2011, the Commission approved the staffs recommendation to develop a staggered submittal and review process for these reviews, and submit a revision to the Enforcement Policy for Commission approval which would propose to extend enforcement discretion to correspond with the new LAR submittal dates. In a letter (ML1 116106160) dated June 10, 2011, the Commission approved the staff's recommendation to publish the Federal Register Notice (FRN) announcing the revision to the Enforcement Policy to extend the enforcement discretion to correspond with a staggered LAR submittal schedule.

On June 23, 2011, SCE&G submitted a letter (RC-11-0099) requesting extension of their enforcement discretion and committed to the submittal date of September 30, 2011. In a letter (RC-11-0161) dated September 30, 2011, SCE&G informed the NRC that additional evaluation and clarification was needed to ensure the Transition Report met the completeness expectation, and that the LAR will be submitted by November 17, 2011.1.1.2.2 Transition Process The transition to NFPA 805 includes the following high level activities: " Elimination of the Self-Induced Station Blackout (SISBO) methodology for prevention of spurious operations of equipment" A new Nuclear Safety Capability Assessment (NSCA) to replace the Appendix R safe shutdown analysis" A new Fire Probabilistic Risk Assessment (Fire PRA) using NUREG/CR 6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, as guidance and a revision to the Internal Events PRAs to support the Fire PRA" A new Non-Power Operations (NPO) assessment" A new Radiological Release assessment" Completion of activities required to transition the pre-transition Licensing Basis to 10 CFR 50.48(c) as specified in NEI 04-02 and RG 1.205 1.2 Purpose The purpose of the Transition Report is as follows: 1) Describe the process implemented to transition the current fire protection program to compliance with the additional requirements of 10 CFR 50.48(c);2) Summarize the results of the transition process;3) Explain the bases for conclusions that the fire protection program complies with 10 CFR 50.48(c) requirements;

4) Describe the new fire protection licensing basis; and 5) Describe the configuration management processes used to manage post-transition changes to the station and the Fire Protection Program, and resulting impact on the Licensing Basis.Introduction Page 3 Introduction Page 3

-ý " RC-11-0149 Section 2.0 2.0 OVERVIEW OF EXISTING FIRE PROTECTION PROGRAM 2.1 Current Fire Protection Licensing Basis Virgil C. Summer Nuclear Station was licensed to operate on August 6, 1982. As a result, the VCSNS fire protection program is based on compliance with 10 CFR 50.48(a), 10 CFR 50.48(b), and the following License Condition:

South Carolina Electric & Gas Company's Virgil C. Summer Nuclear Station license condition 2.c (18) states: Fire Protection System (Section 9.5.1 SSER 4)Virgil C. Summer Nuclear Station shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility, and as approved in the Safety Evaluation Report (SER) dated February 1981 (and Supplements dated January 1982 and August 1982) and Safety Evaluations dated May 22, 1986, November 26, 1986, and July 27, 1987 subject to the following provisions:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of fire.2.2 NRC Acceptance of the Fire Protection Licensing Basis In a letter dated July 25, 1978, the NRC transmitted 21 questions concerning the fire protection evaluation to SCE&G. Responses to these questions were incorporated into the Updated Fire Protection Evaluation Report (FPER) by Revision 2, dated November 30, 1978. Subsequently, the NRC forwarded a second set of 20 questions in a letter dated October 22, 1979. Of particular importance was question 1 of this set which requested that SCE&G demonstrate the capability of the Virgil C. Summer Nuclear Station to achieve safe shutdown for a fire anywhere in the plant including those locations which would require control room evacuation.

Responses to these questions were incorporated into this report by Revisions 4, 5, and 6, dated December 20, 1979, January 1, 1980, and May 15, 1980, respectively.

The NRC reported the results of their evaluation of the original report and the responses to these questions in the original Safety Evaluation Report (SER), dated February 1981, and in Supplemental Safety Evaluation Reports (SSER) 3 and 4, dated January and August 1982, respectively.

During the preparation of the plant license, SCE&G also consented to item 2.c (18) of the operating license which commits SCE&G to maintain the plant fire protection program in accordance with Section Ill.G., III.J., and 111.0. of Appendix R to 10 CFR 50.This action was based on the analysis performed in response to the various questions and on the NRC conclusions in the SER and SSERs.The following accepted deviations previously granted by the NRC in SSER #3 (NUREG-0717, January 1982), including their corresponding Licensing Action ID number, are: 1) Lack of automatic suppression in the Control Room in Fire Area CB-1 7 (LA-CB1 7-011).Overview of Existing Fire Protection Program Page 4 RC-11-0149 Section 2.0 2) Twenty-foot separation not maintained between HVAC chill water pumps in Fire Area IB-07 (LA-IB07-01).

3) Twenty-foot separation not maintained between redundant CC pumps in Fire Area IB-25 (LA-1B25-01).
4) Lack of automatic suppression in the discharge valve rooms and fire detection only in room 25-03 (LA-SWPH05-01).
5) Substantial bullet-proof doors used in lieu of three-hour rated doors in various fire areas (LA-FEAT-04).
6) Back-to-back one and a half-hour rated fire dampers used in lieu of three-hour rated fire dampers in various fire areas (LA-FEAT-05).

The following accepted deviations previously granted by the NRC in SSER # 4 (NUREG-0717, August 1982), including their corresponding Licensing Action ID numbers, are: 1) Lack of automatic suppression in Auxiliary Building rooms ABO1.01.03 85-01, ABO1.07 88-25, ABO1.08.02 97-02, ABO1.04 00-02, ABO1.09, ABO1.10 12-11 North, AB01.18.01 36-18, and ABO1.30 85-01 and Intermediate Building rooms IB10 23-02, IB1l 26-01, IB12 26-02, IB16 51-01, IB17 51-02, IB19 51-03, IB24 36-03B, and IB25.06.01 PA 36-02 (LA-AB01-02, LA-IB10-01, LA-IB11-01, LA-1B12-01, LA-1B16-01, LA-1B17-01, LA-1B19-01, LA-IB24-01

& LA-1B25-05).

2) Lack of automatic fire detection in the areas in Table 9-1 under the column designated

'Deviation Granted by the Staff (LA-YD01-01

& LA-YD02-01).

The following accepted deviation previously granted by the NRC in a Letter to SCE&G, October 1983, including its corresponding Licensing Action ID number, is: 1) Lack of full automatic suppression in Auxiliary Building rooms ABO1.21 (LA-AB01-03).The following accepted deviations/modification previously granted by the NRC in a Letter to SCE&G, May 22, 1986, V.C. Summer Nuclear Station -Appendix R Reanalysis, including their corresponding Licensing Action ID numbers, are: 1) One-hour rate fire barrier not maintained in Fire Area CB-12 (LA-CB12-01).

2) Radiant energy shield used in lieu of a one-hour rated fire barrier in Fire Area IB-25 (LA-1B25-03).
3) Modification

-Eight-hour battery backed emergency lighting not maintained in Fire Area YD-02 (LA-YD02-02).

The following accepted deviations previously granted by the NRC in a Letter to SCE&G, November 26, 1986, V.C. Summer Nuclear Station -Appendix R Reanalysis, including their corresponding Licensing Action ID numbers, are: 1) Three-hour rated fire barrier not maintained between T-hot and T-cold redundant instrument power in Fire Area IB-03 (LA-IB03-01).

2) Three-hour rated fire barrier not maintained between T-hot and T-cold redundant instrument power in Fire Area IB-04 (LA-1B04-01).

Overview of Existing Fire Protection Program Page 5 A- RC-11-0149 Section 2.0 3) Three-hour rated fire barrier not maintained between T-hot and T-cold redundant instrument power in Fire Area IB-25 (LA-1B25-04).

4) Three-hour rated fire barrier not maintained between T-hot and T-cold redundant instrument power in Fire Area RB-01 (LA-RB01-01).

The following accepted deviations previously granted by the NRC in a Letter to SCE&G, July 27, 1987, V.C. Summer Nuclear Station -Appendix R Reanalysis, including their corresponding Licensing Action ID numbers, are: 1) Twenty-foot separation not maintained between chemical volume control cables in Fire Area AB-01 (LA-AB01-01).

2) Three-hour rated fire barrier not maintained between service water system cables in Fire Area MH-02 (LA-MH02-01).
3) One-hour rated fire barrier not maintained between service water booster pump equipment and cables in Fire Area IB-25 (LA-1B25-02).

The following accepted deviations previously granted by the NRC in a Letter to SCE&G, October 10, 1997, Deviation from 10 CFR Part 50. Appendix R,Section III.G. Fire Protection of Safe Shutdown Capability for Virgil C. Summer Nuclear Station, including their corresponding Licensing Action ID numbers, are: 1) Use of one-hour rated Rockbestos Firezone R fire resistant cables in lieu of one-hour rated wrap (LA-CB02-01).

2) Use of one-hour rated Rockbestos Firezone R fire resistant cables in Cable Tray 3088 in lieu of one-hour rated fire barrier (LA-1B25-06).

Overview of Existing Fire Protection Program Page 6 Overview of Existing Fire Protection Program Page 6

.-0 RC-11-0149 Section 3.0 3.0 TRANSITION PROCESS 3.1 Background Section 4.0 of NEI 04-02 describes the process for transitioning from compliance with the current fire protection licensing basis to the new requirements of 10 CFR 50.48(c).NEI 04-02 contains the following steps: " Licensee determination to transition the licensing basis and devote the necessary resources to it;" Submit a Letter of Intent to the NRC stating the licensee's intention to transition the licensing basis in accordance with a tentative schedule;" Conduct the transition process to determine the extent to which the current fire protection licensing basis supports compliance with the new requirements and the extent to which additional analyses, plant and program changes, and alternative methods and analytical approaches are needed;" Submita LAR;" Complete transition activities that can be completed prior to the receipt of the License Amendment;" Receive a Safety Evaluation; and" Complete implementation of the new licensing basis, including completion of modifications identified in Attachment S.3.2 NFPA 805 Process Section 2.2 of NFPA 805 establishes the general process for demonstrating compliance with NFPA 805. This process is illustrated in Figure 3-1. It shows that except for the fundamental fire protection requirements, compliance can be achieved on a fire area basis either by deterministic or RI-PB methods. Consistent with the guidance in NEI 04-02, VCSNS has implemented the NFPA 805 Section 2.2 process by first determining the extent to which its current fire protection program and plant design supports findings of deterministic compliance with the requirements in NFPA 805. RI-PB methods are being applied selectively to the requirements for which deterministic compliance could not be shown.Transition Process Page 7 Transition Process Page 7 RC-11-0149 Section 3.0 Establish fundamental fire NFPA 805 Section 2.2(a)protection elements (Chapter 3)Identify fire hazards NFPA 805 Section 2.2(b)Nuclear safety Identify performance criteria to be Ufe safety NFPA 805 Section 2.2(c) examined Property damagelbusiness (Chapter 1) Interruption Evaluate compliance toI Radiation release performance criteria Identify structures, systems, or components (SSCs) in each fire NFPA 805 Section 2.2(d)area to which the performance criteria applies~NFPA 805 Section 2.2(e) , Deterministic Approach Performance.Based Approach Maintain compliance with existing plant Evaluate ability to satisfy performance license basis (10 CFR 50 App. R, Approved requirements Exemptions, Engineering Evaluations) (Chapter 4)1XZZi 111111 D n aExisting Performance Basis Deterministic Basis Engineering Define fire scenarios and fire design basis Equivalency for each fire area being considered.

Verity deterministic requirements am met Evaluations Evaluate using, e.g., Fire modeling to quantify the fire risk NFPA 805 Section 2.2(f) and margin of safety PSA to examine impact on overall plant risk NFPA 805 Section 2.2(g)Risk-Informed Change EvaluationNFA85Scin2(h

[Evaluate risk impact of changes to FA85Scin2(h the approved design basis Examples Design Basis Documents Fire hazards analysis NFPA 805 Section 2.20) Documentationand configuration Nuclear safety capability assessment Supporting engineering calculations Probabilistic safety analysis Risk-informed change evaluations Feedback 1 ,Establish monitoring program NFPA 805 Section 2.2(i)Figure 3-1 NFPA 805 Process [NEI 04-02 Figure 3-1 based on Figure 2-2 of NFPA 805]2 3.3 NEI 04-02 -NFPA 805 Transition Process NFPA 805 contains technical processes and requirements for a RI-PB fire protection program. NEI 04-02 was developed to provide guidance on the overall process (programmatic, technical, and licensing) for transitioning from a traditional fire protection licensing basis to a new RI-PB method based upon NFPA 805.2 Note: 10 CFR 50.48(c) does not incorporate by reference Life Safety and Plant Damage/Business Interruption goals, objectives and criteria.

See 10 CFR 50.48(c) for specific exceptions to the incorporation by reference of NFPA 805.Transition Process Page 8 Transition Process Page 8

-_6 RC-11-0149 Section 3.0 Section 4.0 of NEI 04-02 describes the detailed process for assessing a fire protection program for compliance with NFPA 805, as shown in Figure 3-2.Transition Report Sect. 4.1 A Transition Report Sect. 4.2 A S fVFDR: Use PB Approach Transition Submit for NRC if Needed (Fire Modeling Report Approval if needed or Fire Risk Evaluations)

Sect. 4.5 Transition Verify I Establish Monitoring Report Program Sect. 4.6 i Transition Confirm I Establish Adequate T Report Documentation I Quality and Ret Configuration Control Sect. 4.7, 5 Regulatory Submittal and Transition Approval IReport Approval__

_ Sect. 4.8, 5 Figure 3-2 Transition Process (Simplified)

[based on NEI 04-02 Figure 4-1 and modified per the VCSNS process]3.4 NFPA 805 Frequently Asked Questions (FAQs)The NRC has worked with NEI and two Pilot Plants (Oconee Nuclear Station and Harris Nuclear Plant) to define the licensing process for transitioning to a new licensing basis under 10 CFR 50.48(c) and NFPA 805. Both the NRC and the industry recognized the need for additional clarifications to the guidance provided in RG 1.205, NEI 04-02, and the requirements of NFPA 805. The NFPA 805 FAQ process was jointly developed by NEI and NRC to facilitate timely clarifications of NRC positions.

This process is described in a letter from the NRC dated July 12, 2006, to NEI (ML061660105) and in Transition Process Page 9 Transition Process Page 9 1VM RC-11-0149 Section 3.0 Regulatory Issues Summary (RIS) 2007-19, Process for Communicating Clarifications of Staff Positions Provided in RG 1.205 Concerning Issues Identified during the Pilot Application of NFPA Standard 805, dated August 20, 2007 (ML071590227).

Under the FAQ Process, transition issues are submitted to the NEI NFPA 805 Task Force for review, and subsequently presented to the NRC during public FAQ meetings.Once the NEI NFPA 805 Task Force and NRC reach agreement, the NRC issues a closure memorandum to indicate that the FAQ is acceptable.

NEI 04-02 will be revised to incorporate the approved FAQs. This is an on-going revision process that will continue through the transition of NFPA 805 non-pilot plants. Final closure of the FAQs will occur when future revisions of RG 1.205, endorsing the related revisions of NEI 04-02, are approved by the NRC. It is expected that additional FAQs will be written and existing FAQs will be revised as plants continue NFPA 805 transition after NRC approval of the Pilot Plant Safety Evaluations.

Attachment H contains the list of approved FAQs not yet incorporated into the endorsed revision of NEI 04-02. These FAQs have been reviewed and/or used to clarify the guidance in RG 1.205, NEI 04-02, and the requirements of NFPA 805 and in the preparation of this LAR.Transition Process Page 10 Transition Process Page 10 A -RC-11-0149 Section 4.0 4.0 COMPLIANCE WITH NFPA 805 REQUIREMENTS

4.1 Fundamental

Fire Protection Program and Design Elements The Fundamental Fire Protection Program and Design Elements are established in Chapter 3 of NFPA 805. Section 4.3.1 of NEI 04-02 provides an industry guideline and systematic process for determining the extent to which the pre-transition licensing basis and plant configuration meets these criteria and for identifying the fire protection program changes that would be necessary for compliance with NFPA 805. SCE&G has developed the Fire Protection Program compliance review with the basic guidance, process and criteria promulgated within these documents.

4.1.1 Overview

of Evaluation Process The comparison of the VCSNS Fire Protection Program to the requirements of NFPA 805 was performed and documented in design calculations that were developed in accordance with the configuration control and quality inherent in SCE&G Design Engineering and as indicated in Section 2.7 of NFPA 805. The NFPA 805 Chapter 3 Code Compliance Document used the guidance contained in NEI 04-02, Section 4.3.1 and Appendix B-1 to develop the documentation package(s) (Reference Figure 4-1).In addition to NFPA 805, Chapter 3 [10 CFR 50.48(c)], applicable codes and/or standards (codes of record) that may have been utilized as design input and are essential to the functional performance of the system and/or feature being reviewed has been incorporated into this overall compliance assessment through the use of individual design calculations that represent NFPA codes that are applicable to VCSNS.The application of these NFPA codes, relate to fire protection systems, structures and features that are considered part of the "power block", as described in NFPA 805 Section 1.6.46, and are "required" systems or features to support the results of the analysis as described in NFPA 805. This includes fire protection systems, structures and features for those plant areas that may be credited to support the nuclear safety performance criteria described in NFPA 805 Section 1.5.1, Nuclear Safety Performance Criteria, or performance based evaluations (see Section 4.8.2 of the Transition Report).Each section and subsection of NFPA 805 Chapter 3 (Table B-i) was reviewed against the current station fire protection program. In some cases multiple compliance statements may have been assigned to a specific NFPA 805 Chapter 3 element.Where this is the case, each compliance basis statement clearly references the corresponding requirement of NFPA 805 Chapter 3. When other NFPA Codes are referenced out of NFPA 805 Chapter 3, the analysis and compliance will be performed and evaluated in a similar manner, as applicable, to the design and features at the station, subject to review and approval by the SCE&G qualified fire protection engineer.Compliance with NFPA 805 Requirements Page 11 Compliance with NFPA 805 Requirements Page I11 RC-11-0149 Section 4.0 Existing~~Secio 4.0-t. ie rtctFm arida~ DeslignrEle~me Ene H In the Compliance Basis H In Reference Document....ld provide toerle ,Field provide:

  • Document References that'Compies
  • No additional demonstrate compliance Document any Open in Compliance Satement clarification Items found duing Review fildieldesoy Eater 'CorEnter u In t Coropliance Basis ield In Refence Document FlPet prvdide: Nis.NoNg 'Coesples nirth Fild prOVide
  • Document References that iEg,°er tgioa t a daoilar on dem.o.strae compliance Docud.....

h Open taFieldntae Cmie nperna ... Enter nprovid C Apocument Fi provd eA. i theo'Comlie 80e as PreVious I Field provide verbatim' Umicnsing Document Rae..-ce Note 1 If the excerpt which provided the tormal approval does not contain suffcient dataIlS of tepremous Chapter 3 Pte o met any Open approval provide an excerpt(s) of *lcenesee subrsentaOs rngardeng the issue for which previous anproval is being inCot Items found durng Review claimed Place te ngce vet of the submetal sobefore the excerpt Of the tonal approvals in the Compiance Basis ,onp Appa...Field...Ia fields he acessaryd Note2 Ene CnnI,., In Comnpliance "MtatFieldt I sefinse, ameundernt il rvdeN Eistits foone dying Eqifaec Evaaie o n peiul nw sGnrcLte 51eautos pos E EiStng Yes of Existing 4Engeenng F provide i Do ment Rerenct that and devsiiog) weme prfonmed for fire p rotecton design varances such as fire s System designs and fire benfier Engineenng Equivalency Evaluation"'

Summaryoftbasesfor demosraecompliance I ouetnyOn component deviations from thespecificfire protection determisthiatrequirements Section 2.2.7oftNFPAa805allows Equiva~wcy in Complince StatementI engneernng evaluationi itesouddngRiw existin EEE that clearly demonstrates an equivalent level of fire protection compared to One deterministic requnirments to Not 3In omiac[Based on elNE I n 0402figren 4-2met ie roie Fg e4 dpc p ce Eun ter prov a nd torrective Accons, as apropenate Note 3:Further Action Rpes led indicates an interim position used du.ng A dtoa de l'Furoher Actiot Required' oFListofAEtsonitinbe lmu References deument any Open the process of completing the B-o Table 'Further Action Required' tnri Do s" fField confi rmiatory activity should be added to the LAR subm ritt el Meet Chapter 3 nor is them Previous Approv*al Choose Enter In Compliance Basis Field In Reference Document Field prvide.0.IOnLcneAmnmn

'License Amendmenlt provide: *Corrective Actions. as atppropriate DcnetayOe CmRequired' Summary of bases for *Document References Itmouend durng Opevie Rqiein ComancFF Statement license amendmentItsfonduigR'e Figure 4-1 Fundamental Fire Protection Program and Design Elements Transition Process[Based on NEI 04-02 Figure 4-213 3 Figure 4-1 depicts the process used as guidance during the transition and therefore contains elements (i.e., open items) that represent interim resolutions.

Additional detail on the transition of FPEEEs is included in Section 4.2.2 of the Transition Report.Compliance with NFPA 805 Requirements Page 12 A-

  • RC-11-0149 Section 4.0 4.1.2 Results of the Evaluation Process 4.1.2.1 Evaluation of NFPA 805 Chapter 3 Requirements The specific requirements for each of the elements in NFPA 805 Chapter 3 are provided in Table B-1 (Attachment A) and provide the results of the technical review performed by VCSNS. The transition of these "Fundamental Fire Protection Program and Design Elements" provides the basis for compliance with the requirements in NFPA 805 Chapter 3. As a result of the activities associated with the review, one or more of the following compliance statement(s) were used: " Complies (C) -The existing FPP elements are determined to meet the requirements of NFPA 805 Chapter 3 element. Acknowledgement and/or restatement of the requirement are not required.

An open item in this category means there are action items to be completed during implementation prior to transition.

Complies directly with the requirements of NFPA 805 Chapter 3." Complies by Alternative (CA) -The existing FPP elements meet the requirements of NFPA 805 by using clarification and/or equivalent alternative(s).

VCSNS requests NRC review/approval of those CA items listed in Section 4.1.2.3 (Table 4-1) of the Transition Report and included in Attachment L.Complies with clarification with the requirements of NFPA 805 Chapter 3." Complies with Fire Protection Engineering Equivalency Evaluations (CE) -The existing FPP elements have been determined to be adequate for the hazard by a FPE and to meet the NFPA 805 Chapter 3 requirements.

Complies through the use of Fire Protection Engineering Equivalency Evaluations (FPEEE) which are valid and of appropriate quality. VCSNS requests NRC review/approval of those Engineering Evaluations listed in Section 4.1.2.3 (Table 4-1) of the Transition Report and included in Attachment L." Complies by Previous NRC Approval (CNRC) -The existing FPP elements specified in NFPA 805 Chapter 3 requirements are not in strict compliance, however, previous NRC approval of the configuration exists. An NRC approved alternative or deviation to NFPA 805 Chapter 3, would supplant the specific requirement of NFPA 805 Chapter 3. Where credited, these prior approvals have been incorporated into an FPEEE, and included in Attachment K and Attachment L." No Review Required (NRR) -The existing Chapter 3 elements are not based on the requirements and/or are not applicable to elements of the VCSNS Fire Protection Program.In addition to these compliance statements, the following approaches were implemented: " An "open item" in any category of the B-1 Table means there are action items to be completed during implementation prior to full transition of the Fire Protection Program. These open items are identified in Attachment S, Table S-2." The use of FPEEEs in the compliance review process is identified by the "CE" designation.

These are utilized to evaluate the requirement and field conditions Compliance with NFPA 805 Requirements Page 13

° RC-11-0149 Section 4.0 and determine the level of compliance and if the element or feature is"Equivalent" or "Adequate for the Hazard." They may also be utilized to assist in the clarification of requirements, past approvals (CNRC) and field conditions that involve complex systems, features or elements that need further understanding within the FPP. The "CE" may be a self-approval, but available for review (see Section 4.2.2 of the Transition Report).The VCSNS process to "self-approve" selected Chapter 3 (Sections

3.8 through

3.11 of NFPA 805) elements and or submit them for NRC review and approval (Sections

3.1 through

3.7 of NFPA 805) is in accordance with the guidance provided in FAQ 06-0008 and RG 1.205 Revision 1.Note: Specific references to VCSNS controlled documents used in this transition report are for reference use only; similar documents, updated revisions or other forms of media may be used to manage this type of information in the future.4.1.2.2 NFPA 805 Chapter 3: "Previous NRC Approval (CNRC)" NFPA 805 Section 3.1 states in part, "Previously approved alternatives from the fundamental protection program attributes of this chapter by the AHJ take precedence over the requirements contained herein." In some cases the prior NRC approval may be unclear for an NFPA 805 Chapter 3 program attribute, to support future clarity for NRC inspections.

In other cases, the requirement has changed from the originally licensed Fire Protection Program attribute.

VCSNS requests that the NRC concur with their finding of prior approval and acceptability for the following sections of NFPA 805 Chapter 3 designated as (CNRC).0 None.4.1.2.3 NFPA 805 Chapter 3: "Compliance Alternatives (CA) Not Previously Approved by NRC" The sections of NFPA 805 Chapter 3 may not have previous NRC approval of an alternate approach, methods and/or condition which VCSNS considers to be minor variations to, and are equivalent to the NFPA 805 requirements.

These "Compliance Alternatives" (CA) are identified in the compliance review of Chapter 3 and satisfy 10 CFR 50.48(c)(2)(vii).

VCSNS requests NRC approval of the proposed alternatives and clarifications of the FPP elements listed in Table 4-1 below. The specific deviation and a discussion of how the alternative satisfies 10 CFR 50.48(c)(2)(vii) requirements are provided in Attachment L.Compliance with NFPA 805 Requirements Page 14 Compliance with NFPA 805 Requirements Page 14 0 0 00 M__ __ ___'11ý-9XWAMM-RC-11-0149 Section 4.0 RC-1I-0149 Section 4.0 Table 4-1 NFPA 805 Chapter 3 Requests for Approval Table B-1 Section Requirement Summary 3.3.1.2 (1) Wood: Clarification and approval is requested for limited use of non-treated wood/lumber for special conditions and operational tasks. Controls are in place to provide the appropriate reviews concerning high risk areas.3.3.5.1 Wiring: Clarification and approval is requested for existing areas of the plant with limited wiring above suspended ceilings that are non-risk significant areas.Engineering Controls exist to mitigate station changes that would require the use of concealed spaces above suspended ceilings for wire routing.3.3.5.3 Electric Cable Construction:

Clarification and approval for existing non-compliant cable and the identified alternative flame propagation tests and controls which may have more rigorous acceptance criteria than IEEE 383-1991.3.3.7.2 Bulk Gas Storage: Clarification and approval is requested for the existing horizontal, hydrogen storage tanks that are perpendicular to the Turbine Building/Control Building based on extensive spatial separation

(>200 feet).3.4.1 (d) Fire Brigade Notification:

Clarification and approval is requested for the verification of a fire by direct visual contact with the fire and/or products of combustion and with direct communication to the control room.3.4.2.4 Pre-Fire Plans: Clarification and approval is requested for the use of multiple procedures to coordinate the fire brigade activities with other groups. The emergency procedures and brigade leader training identifies all support that may originate from the event and require coordination.

3.4.3 (a)(4) Records: Clarification and approval is requested for the use of electronic records and or written records that document fire brigade member training.3.5.15 Yard Fire Hydrant Layout: Clarification and approval is requested regarding the layout of existing yard fire hydrants at the station, considering the requirements found in "approximately every 250 foot spacing" guidance provided by this section.3.6.2 Hose Stations:

Clarification and approval is requested for existing standpipe systems that do not utilize pressure reducers based of fire brigade member hose line training and off-site fire department member training with high pressure hoses.3.6.4 Class III/ Seismic Analyzed Hose Stations:

Clarification and approval is requested regarding the design attributes concerning the existing installation of the Class II Hose Station and Standpipe System at the station.3.8.2 Detection:

Clarification and approval is requested for the existing fire detection layout of devices that are in accordance with NFPA 72E-1 978 code of record.Compliance with NFPA 805 Requirements Page 15 Compliance with NFPA 805 Requirements Page 15 A 'a RC-11-0149 Section 4.0 4.1.3 Definition of Power Block and Plant Where used in NFPA 805 Chapter 3 the terms "Power Block" and "Plant" refer to structures that have equipment required for nuclear plant operations, such as Containment, Auxiliary Building, Service Building, Control Building, Fuel Building, Radioactive Waste, Water Treatment, Turbine Building, and intake structures or structures that are identified in the facility's pre-transition licensing basis.VCSNS reviewed the structures in the Owner Controlled Area to determine those that contain equipment that is required to meet the nuclear safety performance criteria described in Section 1.5 of NFPA 805 and are required for nuclear plant operations.

Note: Structures meeting the radioactive release criteria described in Section 1.5 of NFPA 805, but not required for nuclear plant operations, are separately screened and included in the radioactive release review as discussed in Section 4.4 and Attachment E of the Transition Report.These structures are listed in Attachment I and represent the "power block" and the"plant".4.2 Nuclear Safety Performance Criteria The Nuclear Safety Performance Criteria are established in Section 1.5 of NFPA 805.In addition, Chapter 4 of NFPA 805 provides the methodology to determine the fire protection systems and features required to achieve the performance criteria outlined in Section 1.5. Section 4.3.2 of NEI 04-02 provides a systematic process for determining the extent to which the pre-transition licensing basis meets these criteria and for identifying any necessary fire protection program changes. NEI 04-02, Appendix B-2 provides guidance on documenting the transition of Nuclear Safety Capability Assessment Methodology and the Fire Area compliance strategies.

4.2.1 Nuclear

Safety Capability Assessment Methodology The Nuclear Safety Capability Assessment (NSCA) Methodology review consists of four processes: " Establishing compliance with NFPA 805 Section 2.4.2" Establishing the Safe and Stable Conditions for the Plant" Defining Recovery Actions to be Transitioned" Evaluating Multiple Spurious Operations The methodology for demonstrating reasonable assurance that a fire during non-power operational (NPO) modes will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition is an additional requirement of 10 CFR 50.48(c) and is addressed in Section 4.3 of the Transition Report.4.2.1.1 Compliance with NFPA 805 NSCA (Section 2.4.2)Overview of Process NFPA 805 Section 2.4.2 Nuclear Safety Capability Assessment states: Compliance with NFPA 805 Requirements Page 16 Compliance with NFPA 805 Requirements Page 16 RC-11-0149 Section 4.0"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" The NSCA methodology review evaluated the existing NSCA methodology against the guidance provided in NEI 00-01, Revision 1 Chapter 3, "Deterministic Methodology," as discussed in Appendix B-2 of NEI 04-02. The methodology in Figure 4-2 was used as input in assessing the existing shutdown strategy and consisted of the following activities: " Core methodology documents and plant specific calculations/analyses were gathered." Each specific section of NFPA 805 2.4.2 was correlated to the corresponding section of Chapter 3 of NEI 00-01 Revision 1. Based upon the content of the NEI 00-01 methodology statements, a determination was made of the applicability of the section to the station. The results of the applicability review were then documented in VCSNS Technical Report TR08620-014, "Nuclear Safety Performance Criteria Review Transition Report."" The plant-specific methodology was compared to applicable sections of NEI 00-01 and one of the following alignment statements and its associated basis were assigned to the section: o Aligns o Aligns with intent o Not in Alignment o Not in Alignment, but Prior NRC Approval o Not in Alignment, but no adverse consequences" For the existing shutdown analysis, no additional work was performed relative to corrective actions." Based on the VCSNS approach to transition the station shutdown analysis (FAQ 09-0057), project instructions and guidance were developed to analyze station circuits and documentation in accordance with NEI 00-01. A compliance review with appropriate references is contained in NEI 04-02 Table B-2 (Attachment B) to support the VCSNS transition to NFPA 805.Note: Since NEI 00-01 is a guidance document, portions of its text could be interpreted as 'good practice' or intended as an example of an efficient means of performing the analyses.The comparison of the VCSNS original safe shutdown methodology to NEI 00-01 Chapter 3 (NEI 04-02 Table B-2) was performed and documented in VCSNS Technical Compliance with NFPA 805 Requirements Page 17 Aýýo RC-11-0149 Section 4.0 Report TR08620-014.

Based on the selection of a new shutdown strategy as part of the transition to NFPA 805, including selection of equipment and circuits, NEI 00-01 has been incorporated into the NSCA for VCSNS (Attachment B) to define areas of improvement and support resolution of deficiencies.

Results from Evaluation Process The method used to perform the NSCA with respect to selection of systems and equipment, selection of cables, and identification of the location of equipment and cables, either meets the NRC endorsed guidance directly or met the intent of the endorsed guidance with adequate justification as documented in Attachment B. Table 4-2 includes the analysis criteria that were previously not in alignment for the existing shutdown strategy, and the corresponding improvements that were accomplished during the VCSNS NFPA 805 transition project.Table 4-2 NEI 04-02 Improvements Pre-Transition Assessment Post Transition Alignment (NEI 00-01 Section)The VCSNS Appendix R safe shutdown analysis does The VCSNS NFPA 805 Transition Project has taken into not include comprehensive review/ discussion of fire consideration instrument tubing as a failure mechanism damage to instrument tubing and its impact on for the instrument function.

The tubing is analyzed in instruments credited for plant parameter monitoring, support of instrument operation, documented in a (3.2.1.7 Instrument Tubing) manner similar to "required" cable in the Fire Area Analysis of the Nuclear Safety Capability Assessment.

A line entry for each item in the Composite Equipment A detailed, comprehensive analysis of equipment List indicates the scenario (Compliance Review and/or dependencies has been modeled and included as a part Normal Control Review) for which the item is required, of the NFPA 805 Nuclear Safety Equipment List (NSEL)and separate line items for the support and development in the NFPA 805 Transition Project.supplemental equipment required for the item to function.

However, due to credit of Operator Manual Actions and SISBO, many equipment dependencies were not considered.

(3.2.2.5 Identify Dependencies Between Equipment, Supporting Equipment, Safe Shutdown Systems and Safe Shutdown Paths.)Not all Appendix R safe shutdown equipment had their For the electrical functions/equipment identified in the cables identified, analyzed for circuit failure NFPA 805 NSEL, "required" circuits and circuit failure consequence and located based on the shutdown consequences were evaluated to support the NSCA strategy employed at VCSNS. However, due to credit functions have been targeted, analyzed and of Operator Manual Actions and SISBO, specific incorporated as input files into the Nuclear Safety identification of cables and circuit failure Capability Assessment.

As appropriate, entries will be consequences were not considered.

made into the Corrective Action Program as a part of (3.3.1.1 Cable Selection)

NFPA 805 implementation.

VCSNS Appendix R Analysis does not discuss The NFPA 805 Transition has included electrical devices electrical devices such as relays, switches and signal such as relays, switches and signal resistor units as resistor units as being used as isolation devices, acceptable isolation devices, including devices in (3.3.1.3 Isolation devices) instrument loops.The VCSNS Appendix R safe shutdown analysis The consequences of multiple spurious ESFAS signals evaluates the impact of a single ESFAS actuation.

was evaluated in the NFPA 805 Transition Project to However, the effect of multiple ESFAS actuation was ensure that, although the components may move to their considered to be beyond the expectations of Appendix safe shutdown position, plant transient effects were R requirements.

evaluated.

The logic for MSO scenarios, including impacts of failures have been incorporated NSCA and Compliance with NFPA 805 Requirements Page 18 RC-11-0149 Section 4.0 Table 4-2 NEI 04-02 Improvements Pre-Transition Assessment (NEI00-0 Secion)Post Transition Alignment (NEI 00-01 Section)(3.3.1.6 ESFAS Actuation)

Fire PRA Models.The possible consequences of the spurious operation of certain valves require that the spurious operation be prevented or corrected on a priority basis. The mis alignment results from a change in regulatory philosophy resulting from MSOs. The new philosophy is defined in NEI 00-01 Revision 2; however, for NFPA 805 transition the methodology in FAQ 07-0038 was utilized.

Of immediate concern are the Reactor Coolant System Hi-Lo pressure boundary valves, other valves which can result in loss of reactor coolant inventory, and valves which can result in uncontrolled steam dumping. For other valves, more time is available for correction of spurious operation.

The Reactor Coolant System Hi-Lo pressure boundary valves are all 480V AC motor operated and cannot spuriously open, since power to the motors has been disconnected during normal plant operation.

The remainder of the valves for which spurious operation must be corrected on a priority basis are air operated and controlled by one or more solenoid valves. For valves controlled by a single solenoid, the valve power is disconnected and the cabling to the solenoid is protected with a grounded shield (armor or conduit).This is necessary to prevent a "hot short" from spuriously operating these valves.Valves controlled by 2 or more solenoids, where de-energizing any 1 solenoid puts the valve in the safe position, only require that the valve power be disconnected.

The cabling to the solenoids does not require shielding, since 2 or more "hot shorts" simultaneously would be required to spuriously operate the valve; and for non-reactor coolant pressure boundary valves, multiple hot shorts are not considered credible.

Similarly, for situations involving 2 normally closed valves in series (with individual solenoids) where at least one must be kept closed, disconnecting the power to the solenoids is sufficient.

Two (2) hot shorts, 1 to each solenoid, would be required to cause the flow path to be opened. The required power disconnection was accomplished in the main control board through disconnect switches.A human factors review ensured that the switches can be easily identified.

In addition, a secondary means of disconnecting the solenoid power was provided in a separate fire area, for use in the unlikely event a fire occurs in the Control Room and requires immediate evacuation.

This secondary means of disconnecting power consists of switches in the termination cabinets located in the cable spreading room, which is a separate fire area from the main Control Room.Human factors are also considered in the design of these switches to ensure that they can be readily located and opened. For motor operated valves where sufficient time is available, sDurious operation In the NFPA 805 transition project, the circuit analysis for equipment that were evaluated for spurious operation concerns conformed to the NEI 00-01 methodology that requires that multiple hot shorts (including 3-phase and proper polarity for HLP equipment) will need to be considered for safe shutdown equipment.

In addition, multiple hot shorts were evaluated to determine if a spurious operation would occur to due to multiple hot shorts. Multiple hot shorts causing a single spurious operation, and resolution of multiple spurious operation (MSO) as described in FAQ 07-0038 was addressed in the NFPA 805 Transition.

The NFPA 805 transition has re-evaluated the compliance strategy for mitigating spurious operation on an area-by-area basis. Manual actions that are not allowed by NRC regulations (RIS 2006-10) and that are deemed to be necessary for NSCA compliance have been evaluated as part of a Fire Risk Evaluation as acceptable "recovery actions." In general, recovery actions have been minimized for the VCSNS NFPA 805 Transition.

The methodology for mitigating spurious operation also credits opening disconnect switches in the Control Room (or outside the Control Room when Control Room evacuation is necessary).

A review of the electrical schematic for these solenoid valves to determine that the disconnect switches affect both the positive and negative legs of the circuitry was specifically addressed as part of the NFPA 805 Transition.

Comliace ithNFP 80 RqIreetsPge1 Compliance with N1FPA 805 Requirements Page 19 RC-11-0149 Section 4.0 Table 4-2 NEI 04-02 Improvements Pre-Transition Assessment Post Transition Alignment (NEI 00-01 Section)was controlled by opening the cubicle breaker in the MCC and then the valve manually repositioned locally by operating the hand wheel. The Fire Emergency Procedures

[FEP] direct the tripping of the MCC cubicle breakers in a timely manner and provide separate instructions to manually reposition valves for which spurious repositioning could be detrimental to safe shutdown.

These actions were consistent with the Appendix R Shutdown strategy.

In many cases, manual actions at defined locations precluded spurious operation of the potentially affected equipment.

For evaluation of spurious operation of equipment, the methodology that "The cabling to the solenoids does not require shielding, since 2 or more "hot shorts" simultaneously would be required to spuriously operate the valve; and for non-reactor coolant pressure boundary valves, multiple hot shorts are not considered credible" does not conform to the NEI 00-01 methodology that requires that multiple hot shorts (3-phase and proper polarity) will need to be considered for HLP equipment and safe shutdown equipment.

(3.3.2 A Cables Whose Failure May Cause Spurious Actuations)

The method of the Appendix R safe shutdown analysis was to first identify the source power circuit breakers and their "associated" breakers.

After that determination, a simple one-line diagram was prepared identifying the frame size and trip setting of each breaker. From this data and the manufacturer circuit breaker trip characteristic curves, coordination curves were prepared to demonstrate visually the amount of coordination existing between the associated circuit breakers.

A complete report of this coordination study was prepared and made part of the Appendix R review documentation.

The results of the analysis indicated a high degree of coordination between the protective devices for the associated circuits of interest and the main protective devices for required power sources. Several cases for which the degree of coordination was insufficient were identified, and suitable new trip setting values for the circuit breakers were established and implemented.

This review demonstrated that the circuit breakers were coordinated in accordance with accepted design practices and that required power sources will be adequately protected from fire induced faults on circuits "associated by common power supply." (3.3.2 B Common Power Source Cables)The NFPA 805 transition project has analyzed common power supplies required to be energized for the NSCA function to ensure compliance with NEI 00-01. Cases where breaker coordination has been determined to be insufficient, entries will be made into the Corrective Action Program as a part of NFPA 805 implementation.

An electrical circuit term sheet is issued for each electrical cable or circuit. Each circuit sheet provides information to the field for terminating the "from" and"to" ends of each circuit. S-212-001 Sheets 1 -11"Required" circuits, including field routing necessary to support the selected equipment, have been identified and evaluated as part of detailed circuit analysis Dackage for electrically operated NFPA 805 equiDment Compliance with NFPA 805 Requirements Page 20 Compliance with NFPA 805 Requirements Page 20 e RC-11-0149 Section 4.0 Table 4-2 NEI 04-02 Improvements Pre-Transition Assessment Post Transition Alignment (NEI 00-01 Section)provide a complete description of all required fields needed for completing the electrical circuit schedule.An electrical cable pull slip is issued for each electrical cable or circuit. Each pull slip provides information to the field for routing circuits through raceways.Normally, cable pull slips are developed using the computerized cable management database, PC-CKS.Cable pull slips can be developed manually but, the PC-CKS Database should be used to verify that criteria such as raceway separation, combustible fire loading, percent fill, and weight loading are maintained.

For PC-CKS to route or verify the routing of a circuit, any new cable bill of material (B/M) or new multi-cable conduit (XX) must be entered first. Refer to the following attachments:

  • Not all Appendix R safe shutdown equipment had their cables identified, analyzed for circuit failure consequence and located. For all equipment credited for Nuclear Safety Performance (NSP), the cable selection will need to be determined, analyzed and located. These have been entered into the Corrective Action Program." PC-CKS includes routing of all cable trays, but not conduits.

Walkdown of conduit locations will need to be performed and entered into PC-CKS.and functions.

The results of the circuit analysis were incorporated into PC-CKS, as necessary to support the NSCA and Fire PRA analysis.To meet this requirement, certain equipment and circuits must remain functional in the event of a fire.ES-427 provides screening and questions related to Appendix R applicability.

If required, additional reviews shall be completed to address affects or involvement of an Appendix R related system (EE-06). Refer to Cable & Raceway DBD section 4.5 for further explanation of Appendix R applicability.

(3.3.3.4 Identify Routing of Cables)The effects of fire damage to instrument tubing were The VCSNS NFPA 805 Transition Project has taken into not documented in the FPER. consideration instrument tubing as a failure mechanism (3.4.1.8 Consider Instrument Tubing Effects) to the instrument function.

The tubing is analyzed in support of instrument operation, documented in a manner similar to "required" cable in the fire area analysis of the NSCA.Compliance with NFPA 805 Requirements Page 21 Compliance with NFPA 805 Requirements Page 21 RC-11-0149 Section 4.0 RC-1 1-0149 Section 4.0 Step I Step 2 Stop 3 -Step 4 Figure 4-2 Summary of Nuclear Safety Methodology Review Process (FAQ 07-0039)4.2.1.2 Safe and Stable Conditions for the Plant Overview of Process The nuclear safety goals, objectives and performance criteria of NFPA 805 allow more flexibility than the previous deterministic programs based on 10 CFR 50 Appendix R and NUREG 0800, Section 9.5-1 (and NEI 00-01, Chapter 3) since NFPA 805 requires the licensee to maintain the fuel in a safe and stable condition rather than achieve and maintain cold shutdown.NFPA 805, Section 1.6.56, defines "Safe and Stable Conditions" as follows: "For fuel in the reactor vessel, head on and tensioned, safe and stable conditions are defined as the ability to maintain 99, with a reactor coolant temperature at or below the requirements for hot shutdown for a boiling water reactor and hot standby for a pressurized water reactor. For all other configurations, safe and stable conditions are defined as maintaining Keff<O. 99 and fuel coolant temperature below boiling." The nuclear safety goal of NFPA 805 requires "...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" without a specific reference to a mission time or event coping duration.For the plant to be in a safe and stable condition, it may not be necessary to perform a transition to cold shutdown as currently required under 10 CFR 50, Appendix R.Therefore, the unit may remain at or below the temperature defined by a hot standby/hot shutdown plant operating state for the event.Compliance with NFPA 805 Requirements Page 22

° RC-11-0149 Section 4.0 Results Based on VCSNS Technical Report TR08620-312, "Nuclear Safety Capability Assessment Report," the NFPA 805 licensing basis for VCSNS is to achieve and maintain Hot Standby, Mode 3, which is the basic safe and stable condition established and maintained for the NSCA. The 5 nuclear safety performance criteria (Reactivity Control, Coolant Inventory, Decay Heat Removal, Process Monitoring, and Vital Auxiliaries) are achieved by existing plant systems. Some systems, such as the Chemical and Volume Control System (CVCS), serve multiple goals of Coolant inventory addition and Boric acid addition for long term reactivity control. Following the initial coping/assessment period at the start of a fire, the operators will maintain safe and stable conditions as follows: Safe and Stable Summary Description Reactivity Control is achieved by control rod insertion by manual Reactor Trip from the main control board. Operating limits on control rod bank positions assure that adequate reactivity insertion will occur, with margin. Boration is needed to maintain shutdown reactivity margin during cooldown -this is provided with the Chemical and Volume Control System (CVCS) system supplying water from the Refueling Water Storage Tank (RWST)." Coolant Inventory is maintained by the CVCS based on a number of parameters including pressurizer level. The post-fire shutdown plan includes isolation of letdown to preserve Reactor Coolant System (RCS) inventory along with throttling charging pump injection flow to avoid RCS overfill.

The assumed flow path is from the RWST (normal Volume Control Tank path is isolated) to the charging pump(s) and into the RCS via either the normal or safety injection path.Charging Pump miniflow is maintained in the open position.

RCS pressure control is maintained by the ability to increase pressure by an emergency bus supplied pressurizer heater bank or by control of the charging rate, and by the ability to reduce pressure by pressurizer PORV operation." An important part of maintaining RCS inventory is maintaining Reactor Coolant Pump (RCP) seal integrity.

RCP seal cooling is maintained by either the charging pump seal injection path or the Component Cooling (CC) flow to the RCP thermal barrier heat exchanger.

Modifications are planned (see Table S-1 in Attachment S) to provide a redundant seal injection system that is independent of the existing system and not affected in the problem fire areas. Second, a new seal material is planned (see Table S-1 in Attachment S) so that the loss of seal cooling does not lead to significant loss of RCS inventory.

Until new seal materials are installed, procedures for seal cooling interruptions are in place to address the issue as a part of the existing appendix R analysis." As part of the NSCA analysis, potential failures to components that affect RCS inventory including Power Operated Relief Valves (PORVs), failure to isolate letdown, charging pump failure, and issues associated with the Reactor Coolant Pump (RCP) seals have been considered and included in the shutdown model.Compliance with NFPA 805 Requirements Page 23 Compliance with NFPA 805 Requirements Page 23 A-° RC-11-0149 Section 4.0 Decay heat removal following reactor trip is provided by Emergency Feedwater System (EFW) to the Steam Generators (SGs) and atmospheric relief of steam through the safety valve(s).

Other systems may be available but were not credited for the deterministic evaluation.

The Thermal Hydraulic (TH) analysis also considered cooldown with only one SG available.

EFW supply is initially from the Condensate Storage Tank (CST) with backup from Service water.Other sources of cooling water are not precluded.

EFW flow control utilizes the flow control valves and includes the ability to isolate a Steam Generator or secure pump(s) as determined by the operator." Instrumentation for the transition to (and maintenance of) Hot Standby (Mode 3)consists of RCS wide range pressure, Pressurizer level, Nuclear Source Range indication, Steam Generator Pressure and level, and RCS temperature (preferably Thot and Tcold from the steaming Steam Generator(s) loops)." Support Systems are required for almost all safety functions and include electrical power, Service Water (SW), CC, Chilled Water, room cooling, containment cooling, and ventilation for specific rooms. Systems typically not credited (but potentially available) include instrument air, secondary side support, Industrial cooling, and other plant systems not associated with a safety function.The electrical system includes switchgear, transformers, inverters, panels, and the diesel generators." If evacuation of the Main Control Room was required due to a significant fire in the Control Complex, the Control Room Evacuation Panel (CREP) is designed to provide the Instrumentation and controls to maintain Hot Standby, as a Primary Control Station.Demonstration of the Nuclear Safety Performance Criteria for safe and stable conditions was performed in two analyses." The At-Power analysis is discussed in Section 4.2.4 of the Transition Report.This analysis, which is initiated in Modes 1-2, includes actions to achieve Hot Standby. In addition, those actions necessary to achieve Cold Shutdown from Hot Standby are described." The Non-Power analysis is discussed in Section 4.3 of the Transition Report.This analysis evaluates Systems and Components for Mode 3 and below.After conditions stabilized, operators can initiate systems required for cooldown and depressurization to achieve and maintain Cold Shutdown.

The ability to cool down to Cold Shutdown (Mode 5) is considered a subset of the NSCA at VCSNS. The compliance review demonstrates the ability to achieve and maintain safe and stable hot standby conditions.

However, in the event the plant decides to transition to cold shutdown conditions, the actions needed to transition from hot standby to cold shutdown are also documented in the NSCA Report.Cooldown Summary Description Mode 3 -The cooldown process begins in Mode 3 and uses the SG PORVs to reduce pressure below the setpoint of the SG safety valves, which in turn cools the RCS. The EFW system flowpath and function stay the same, though slightly more flow may be Compliance with NFPA 805 Requirements Page 24 A -6 RC-11-0149 Section 4.0 needed. Reactivity control consists of adding borated water from the RWST to the RCS-the charging pumps have ample capacity to accommodate the RCS shrinkage.

The same 'safe and stable' flowpath is used -the borated water will assure that shutdown margin is maintained.

Inventory control and decay heat removal uses the same methods as identified in Section 4.3.2 of the Transition Report. Likewise, the 'steady-state' Mode 3 equipment is also used for the 'cooldown' Mode 3 for the Instrumentation and Support Systems. Pressure control again uses the same equipment, but the pressure is controlled to permit blocking Engineered Safeguards Features Actuation Signals (ESFAS) and accumulator discharge to reduce pressure to the Residual Heat Removal (RHR) operating conditions (temperature and pressure).

Recovery Action(s)outside the primary control station(s) may be needed for the transition from Hot Standby to Cold Shutdown.Mode 4 -The transition to Mode 4 (Hot Shutdown) entails a number of steps to prepare the RHR system for connection to the RCS. The 5 nuclear safety performance criteria are met as follows: " Reactivity control requirement entails a Boron concentration measurement for the RHR system" Inventory control concern are the same -control pressure" Decay heat removal adds RHR heat exchanger flow control" Instrumentation adds the RHR instruments" Support systems are the same but involve a different line-up. The RHR suction valves (8701A/B and 8702A/B) have an interlock that -depending on fire damage-may need a Recovery Action to open, and then RHR is available and the plant enters Mode 4. Once in Mode 4, EFW can be turned off and decay heat removal is by the RHR system.Mode 5 -Cooldown to Cold Shutdown (Mode 5, RCS<200 Deg F) uses the same equipment as Mode 4 and may proceed without further significant recovery actions.Other operational concerns include mode-dependent ESF equipment operability and equipment racked out for overpressure concerns.The ability to achieve Cold Shutdown, including any necessary cooldown actions, is documented in VCSNS Technical Report TR08620-312 on a Fire Area basis.4.2.1.3 Establishing Recovery Actions Overview of Process NEI 04-02 and RG 1.205 suggest that a licensee submit a summary of its approach for addressing the transition of Operator Manual Actions (OMA) as recovery actions in the LAR (Regulatory Position 2.21 and NEI 04-02, Section 4.6). As a minimum, NEI 04-02 suggests that the assumptions, criteria, methodology, and overall results be included for the NRC to determine the acceptability of the licensee's methodology.

The discussion below provides the methodology used to define and assess the Recovery Actions necessary to support the goals of the NFPA 805 Nuclear Safety Compliance with NFPA 805 Requirements Page 25 Compliance with NFPA 805 Requirements Page 25 A e RC-11-0149 Section 4.0 Capability Assessment for VCSNS. This process was initially based on FAQ 07-0030 (ML1 10070485) and consists of the following steps: " Step 1: Define the primary control station(s) and determine which pre-transition OMAs are taken at the primary control station(s).

Note: Activities that take place at primary control station(s), including those required to enable the primary control station, or in the Main Control Room, are not recovery actions, by definition (Reg Guide 1.205, Section 2.4)." Step 2: Determine the population of recovery actions that are required to resolve VFDRs, and are therefore subject to a risk informed evaluation (including defense in depth considerations)." Step 3: Evaluate the additional risk of the use of recovery actions." Step 4: Evaluate the feasibility of the recovery actions." Step 5: Evaluate the reliability of recovery actions.Results The population of Recovery Actions credited for compliance with NFPA 805 is included in Attachment G, Table G-1. The risk associated with the Recovery Actions, including an assessment of Feasibility and Reliability, are documented in, "Fire PRA Human Reliability Analysis Report," which is found in VCSNS Design Calculation DC00340-001,"Fire PRA Plant Final Report," Attachment 10 and "Fire Risk Evaluation Report NFPA 805," PRA Evaluation 11-04. Table G-2 provides the bounding delta Human Error Probability (HEP) calculation that was used to model the elimination of the use of recovery actions and obtain the additional risk of recovery.4.2.1.4 Evaluation of Multiple Spurious Operations Overview of Process The prevailing guidance for consideration of Multiple Spurious Operations (MSOs) is provided in the FAQ 07-0038 closeout memorandum dated February 3, 2011 (ML110140242).

As part of the NFPA 805 transition project, a review and evaluation of VCSNS susceptibility to fire-induced MSOs was performed.

The original process was conducted in accordance with NEI 04-02, Revision 2 and RG 1.205, and was supplemented by FAQ 07-0038 Revision 3 as the review progressed.

The original approach outlined in Figure 4-3 (based on Figure 4-8 from FAQ 07-0038 Rev 1) is similar to the method to address fire-induced MSOs in the final process approved in FAQ 07-0038 Rev 3. The method to support the transition to NFPA 805 was refined to consist of the following steps: " Identify potential MSOs of concern, based on Draft E PWROG Generic MSO list dated March 26, 2008." Conduct an expert panel to assess plant specific vulnerabilities (e.g., per NEI 00-01, Rev. 1 Section F.4.2)." Update the Fire PRA model and NSCA to include the MSOs of concern." Evaluate for NFPA 805 compliance." Document results.Compliance with NFPA 805 Requirements Page 26 RC-11-0149 Section 4.0 The process and inputs are described in VCSNS Technical Report TR08620-025.

The results are integrated into the Fire PRA and NSCA models and support the transition to a new licensing basis. The Post-transition assessment of a specific MSO would be a simplified version of this process, and may not need the level of detail shown in the following section (e.g., an expert panel may not be necessary to identify and assess a new potential MSO). Identification of new potential MSOs will be part of the plant change review process, Industry OE Review and/or Self-Assessment process.Compliance with NFPA 805 Requirements Page 27 Compliance with NFPA 805 Requirements Page 27 a Of -- -RC-11-0149 Section 4.0 Step 1 Step 2 Step 3 Step 4 Step 5 Step 6 Step 7 Document Results Figure 4-3 Multiple Spurious Operations

-Transition Resolution Process (Based on FAQ 07-0038 Revision I and later modified)Compliance with NFPA 805 Requirements Page 28 Compliance with NFPA 805 Requirements Page 28 A RC-11-0149 Section 4.0 Results Refer to Attachment F for the process used by VCSNS and the results from the process.4.2.2 Fire Protection Engineering Equivalency Evaluation (FPEEE) Transition Overview of Evaluation Process The FPEEEs that support compliance with NFPA 805 Chapter 3 or Chapter 4 (both those that existed prior to the transition and those that were created during the transition) were reviewed using the methodology contained in NEI 04-02. The methodology for performing the FPEEE review includes 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 accordance with the guidance in RG 1.205, Regulatory Position 2.3.2, and NEI 04-02, as clarified by FAQ 07-0054, Demonstrating Compliance with Chapter 4 of NFPA 805, FPEEEs that demonstrate that a fire protection system or feature is "adequate for the hazard" are summarized 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 K or Attachment L, as appropriate for NRC review and approval.When NRC approval is requested or required, the reliance on FPEEEs to demonstrate compliance with NFPA 805 requirements was documented in the LAR.Results The review results for FPEEEs are documented in SCE&G controlled documents, and summarized in Table 4-3 in Section 4.2.3 of the Transition Report. Any FPEEEs where SCE&G request specific NRC review and approval are included in Attachment K or Attachment L. Other FPEEEs, not submitted for NRC approval are controlled, and available for onsite review.Compliance with NFPA 805 Requirements Page 29 Compliance with NFPA 805 Requirements Page 29

@- RC-11-0149 A ý C-Section 4.0 4.2.3 Licensing Actions: Resulting NFPA 805 Analysis & Transitions Overview of Evaluation Process The review of new and/or existing licensing actions (deviations) was performed in accordance with NEI 04-02. The methodology for the licensing action review included the following: " Determination of the bases for acceptability of the licensing action." Determination that these bases for acceptability are still valid and required for NFPA 805." Incorporation of existing, credited licensing actions into FPEEEs.Results* As a result of the review, selected Deterministic Requirement Open Item Descriptions (DROIDs) (see Attachment C) were identified that were deterministic (NFPA 805, Chapter 4.2.3) in nature and were dispositioned with prior NRC approval or requires NRC approval.

These actions are summarized below in Table 4-3, documented in FPEEEs, with the licensing action itself, and if previously existing, are identified in Attachment K.* When identified, the previous licensing actions will be transitioned into the NFPA 805 fire protection program as previously approved per NFPA Section 2.2.7 or as new licensing actions requiring approval per NFPA Section 4.2.3. Upon approval, these licensing actions are considered compliant under 10 CFR 50.48(c).Table 4-3 NSCA FPEEs/ Licensing Actions FPEEE Licensing Action Description Auxiliary Building:

Assess the lack of 20 foot separation and TR0780E-001, ABO0-01 LA-AB01-01 full automatic suppression in fire zone ABO1.09 for compliance with Section 4.2.3.3(c) of NFPA 805-2001.Auxiliary Building:

Assess the lack of automatic fire TR0780E-001, AB01-02 None suppression in fire zone ABO1.08 for compliance with Section 4.2.3.3(b) of NFPA 805-2001.Auxiliary Building:

Assess the lack of full automatic TR0780E-001, AB01-03 LA-AB01-03 suppression in fire zone ABO1.21 for compliance with Section 4.2.3.3(b) of NFPA 805-2001.Auxiliary Building:

Assess the adequacy of specific barriers that have been credited in the Nuclear Safety Capability TR0780E-001, AB01-04 None Assessment (NSCA) as a feature that provides sufficient fire resistive capability to prevent fire damage outside the fire area/zone per Section 4.2.3.3(a) of NFPA 805-2001.Control, Intermediate Buildings:

Assess the use of Rockbestos Firezone R cable as a replacement for a 1-hour TRO780E-001, CB25-01 LA-IB25-06 barrier for redundant safe shutdown equipment/circuits in Fire Area CB02 and fire zone IB25.01.02 to comply with Section 4.2.3.3(c) of NFPA 805-2001.Electrical Underground Duct Bank: Assess the lack of 20 TRO780E-001, DB-01 None foot separation between Train "A" and Train "B" circuits in some of the Electrical Duct Banks (Fire Area DB), a lack of automatic fire detection and a lack of automatic fire suppression in Fire Compliance with NFPA 805 Requirements Page 30 RC-11-0149 Section 4.0 Table 4-3 NSCA FPEEs/ Licensing Actions FPEEE Licensing Action Description Area DB in order to comply with Section 4.2.3.3(b) of NFPA 805-2001.Fuel Handling Building:

Assess the lack of full automatic fire TR0780E-001, FH01-01 None suppression throughout fire area FHO1 in the Fuel Handling Building in order to comply with Section 4.2.3.3(b) of NFPA 805-2001.Intermediate Building:

Assess the lack of 20 foot separation in TR0780E-001, IB07-01 LA-IB07-01 fire area IB07 for compliance with Section 4.2.3.3(b) of NFPA 805-2001.Intermediate Building:

Assess the lack of 20 foot separation in TR0780E-001, IB25-02 LA-IB25-01 fire zone 1125.01 for compliance with Section 4.2.3.3(b) of NFPA 805-2001.Intermediate Building:

Assess the lack of 20 feet of physical TR0780E-001, IB25-03 LA-IB25-03 separation between fire subzones 1B25.01.01 and 1125.01.02 for compliance with Section 4.2.3.3(b) of NFPA 805-2001.East Penetration Access Area: Assess the lack of automatic TRO780E-001, IB25-04 None fire suppression in fire zone IB25.03 East Penetration Access Area (Room 12-01) in order to comply with Section 4.2.3.3(b) of NFPA 805-2001.Intermediate, Reactor Buildings:

Assess the continuous TR0780E-001, IB25-05 / LA-IB25-04 availability of process monitoring equipment in fire zones RB01-01 LA-RB01-01 RB01.01.01 and IB25.04 in order to comply with Section 1.5.1(a), (b), (c), and (d) and Section 4.2.3.3(a) or Section 4.2.3.4(b) of NFPA 805-2001.Intermediate Building:

Assess the lack of automatic fire TRO780E-001, IB25-06 LA-IB25-05 suppression in Fire Area IB25, specifically fire subzone IB25.06.02 containing redundant safe shutdown circuits in order to comply with Section 4.2.3.3(b) of NFPA 805-2001.Intermediate Building:

Assess the adequacy of specific barriers that have been credited in the Nuclear Safety TR0780E-001, IB25-07 None Capability Assessment (NSCA) as a feature that provides sufficient fire resistive capability to prevent fire damage outside the fire area/zone per Section 4.2.3.3(a) of NFPA 805-2001.Electrical Man Hole: Assess the lack of 20'-0" of physical separation between Train "A" and Train "B" equipment/circuits, TR0780E-001, MH02-01 LA-MH02-01 the lack of automatic fire detection, and the lack of automatic fire suppression in fire area MH02 for compliance with Section 4.2.3.3(b) of NFPA 805-2001.Service Water Pump House: Assess the lack of 20 foot of TR0780E-001, SWPH05-01 LA-SWPH05-01 separation in fire area SWPH05 for compliance with Section 4.2.3.3(b) of NFPA 805-2001.Yard, CST: Assess the lack of automatic fire suppression and TR0780E-001, YD02-01 LA-YD02-01 fire detection in fire zones YD02.01 and YD02.02 for compliance with Section 4.2.3.3(b) of NFPA 805-2001.Various Areas: Determine the adequacy of the fire resistive TRO780E-006, FEAT-02 None capabilities for the embedded electrical raceway conduits and their associated pull boxes containing cables/circuits important to safety and safe shutdown of the plant.Various Areas: Review and document previous NRC approval TR0780E-006, FEAT-04 LA-FEAT-04 for the installation of unlisted fire doors in select fire barriers required per the performance requirements established by Compliance with NFPA 805 Requirements Page 31 A- e RC-11-0149 Section 4.0 Table 4-3 NSCA FPEEsl Licensing Actions FPEEE Licensing Action Description National Fire Protection Association NFPA 805-2001, Chapter 4 Section 3.11.Various Areas: Review and document previous NRC approval for the installation of dual 11/2 hour rated fire dampers mounted TR0780E-006, FEAT-05 LA-FEAT-05 "back-to-back" in 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rated fire barriers required per the performance requirements established by NPFA 805-2001, Chapter 4 and Section 3.11.The licensing actions listed in Table 4-4 are no longer necessary and will not be transitioned into the NFPA 805 fire protection program.Table 4-4 Licensing Actions Not Being Transitioned to NFPA 805 Licensing Description Reason No Longer Necessary Action ID LA-AB01-02 Appendix R Deviation, Various Areas -Lack No compliance strategy utilized in this area for of Automatic Suppression (III.G.2 criteria)

NFPA 805 requires automatic suppression.

LA-CB12-01 Appendix R Deviation, Control Building -Circuits for IN100031 no longer routed in CB12. 1-Lack of 1-hour fire rated barrier (llI.G.2.c hr barrier no longer required.criteria)LA-CB17-01 Appendix R Deviation, Control Building -Performance-Based methods in NFPA 805 Lack of Automatic Suppression (III.G.3 include analysis for lack of suppression in the criteria) control room. Therefore, this Licensing Action will not be transitioned into NFPA 805.LA-IB03-01 Appendix R Deviation, Intermediate Building -All RCS Temperature for indication at the MCB is Lack of 3-hour fire rated barrier (IIl.G.2.a embedded in 1B03. Embedded conduits criteria) evaluated in TR0780E-001.

LA-IB04-01 Appendix R Deviation, Intermediate Building -Current NSCA model does not reveal a need to Lack of 3-hour fire rated barrier (llI.G.2.a credit substitution of Process Monitoring criteria) instruments.

LA-IB10-01 Appendix R Deviation, Various Areas -Lack Fire Risk Evaluation utilized to address lack of of Automatic Suppression (III.G.2 criteria) automatic suppression in the area.LA-IB1 1-01 Appendix R Deviation, Various Areas -Lack Utilizing PB Fire Modeling in this area to resolve of Automatic Suppression (III.G.2 criteria) lack of automatic suppression failure.LA-IB12-01 Appendix R Deviation, Various Areas -Lack Fire Risk Evaluation utilized to address lack of of Automatic Suppression (III.G.2 criteria) automatic suppression in the area.LA-IB16-01 Appendix R Deviation, Various Areas -Lack No compliance strategy utilized in this area for of Automatic Suppression (III.G.2 criteria)

NFPA 805 requires automatic suppression.

LA-IB17-01 Appendix R Deviation, Various Areas -Lack No compliance strategy utilized in this area for of Automatic Suppression (III.G.2 criteria)

NFPA 805 requires automatic suppression.

LA-IB19-01 Appendix R Deviation, Various Areas -Lack No compliance strategy utilized in this area for of Automatic Suppression (III.G.2 criteria)

NFPA 805 requires automatic suppression.

LA-IB24-01 Appendix R Deviation, Various Areas -Lack No compliance strategy utilized in this area for of Automatic Suppression (III.G.2 criteria)

NFPA 805 requires automatic suppression.

LA-YD01-01 Appendix R Deviation, Various Areas -Lack Fire Risk Evaluation utilized to address lack of of Automatic Fire Detection (III.F criteria) automatic suppression in the area.LA-YD02-02 Appendix R Modification, Yard Areas -Lack Human Reliability Analysis includes factors such Compliance with NFPA 805 Requirements Page 32 RC-11-0149 Section 4.0 Table 4-4 Licensing Actions Not Being Transitioned to NFPA 805 Licensing Description Reason No Longer Necessary Action ID of 8-hr battery backed emergency lighting as lack of emergency lighting in NFPA 805.(lll.J criteria)

Licensing action for lack of Emergency Lighting not required to be transitioned to NFPA 805.VCSNS was licensed to operate after January 1, 1979 and as such, 10 CFR 50 Appendix R is not applicable and exemptions from the regulation were not necessary.

Since the deviations are either compliant with 10 CFR 50.48(c) or no longer necessary, as discussed in Attachment M, upon issuance of the new 10 CFR 50.48(c) license condition, the current VCSNS license condition will be superseded.

VCSNS 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.4.2.4 Fire Area Disposition Overview of Evaluation Process The Fire Area Transition (NEI 04-02 Table B-3) was performed using the methodology contained NEI 04-02, FAQ 07-0054 and FAQ 09-0057. The methodology for performing the Fire Area Transition, depicted in Figure 4-4, is outlined below.Note: Throughout this report, deterministic open items are referred to as DROIDs.DROIDs that are dispositioned using performance-based evaluations (fire modeling or fire risk evaluations) are referred to as Variances From the Deterministic Requirements (VFDRs).Step 1 -Assembled documentation.

Gathered industry and plant-specific fire area analyses and licensing basis documents.

Step 2 -Documented fulfillment of nuclear safety performance criteria." Assess accomplishment of nuclear safety performance goals. Documented the method of accomplishment, in summary level form, for the fire area. The overview of accomplishment of each of the performance goals is included in Attachment C." Documented evaluation of effects of fire suppression activities.

Documented the evaluation of the effects of fire suppression activities on the ability to achieve the nuclear safety performance criteria." Performed licensing action reviews. Performed a review of the licensing aspects of the selected fire area and document the results of the review. See Section 4.2.3 of the Transition Report." Performed fire protection engineering equivalency evaluation reviews.Performed a review of fire protection engineering equivalency evaluations (or created new evaluations) documenting the basis for acceptability.

See Section 4.2.2 of the Transition Report.Compliance with NFPA 805 Requirements Page 33 Compliance with NFPA 805 Requirements Page 33 A-e RC-11-0149 Section 4.0" Defined recovery actions to support NSCA performance goals to determine those actions taking place outside of the main control room or outside of the primary control station(s).

See Section 4.2.1.3 of the Transition Report." Defined modifications to achieve deterministic compliance with the NSCA criteria on a case by case basis.Step 3 -DROID Identification and characterization.

For those items in Step 2 that were not resolved by deterministic compliance (NFPA 805, Section 4.2.3), the process identified DROIDs. Developed DROID problem statements to support resolution.

Step 4 -Preliminary Disposition.

Define options to disposition DROIDs, which may include modifications or performance-based evaluations (fire modeling or fire risk evaluations).

For resolution via performance-based evaluations (VFDRs), see Section 4.5.2 of the Transition Report for additional information.

Step 5 -Final Disposition." Documented final disposition of the DROIDs and, as applicable, VFDRs in Attachment C (NEI 04-02 Table B-3)." For Recovery Action compliance strategies, ensured the HRA of the required recovery actions was completed.

See Section 4.2.1.3 of the Transition Report for additional information." Documented the post transition NFPA 805 Chapter 4 compliance basis in Attachment C.Step 6 -Documented

'Required' fire protection systems and features.

Reviewed the NFPA 805 Section 4.2.3 compliance strategies (including fire area licensing actions and engineering evaluations) and the NFPA 805 Section 4.2.4 compliance strategies (including simplifying deterministic assumptions) to determine the scope of fire protection systems and features 'Required' by NFPA 805 Chapter 4. The 'Required' fire protection systems and features are subject to the applicable requirements of NFPA 805 Chapter 3. For additional discussion, see Section 4.8.2 of the Transition Report.Compliance with NFPA 805 Requirements Page 34 Compliance with NFPA 805 Requirements Page 34 Aý*AoN-,A,-

RC-11-0149 Section 4.0 Document Final Disposition of DROID Compliance options include: Accept As Is Require FP systems/features Require Recovery Action Require Programmatic Enhancements Require Plant Modifications (NEI 04-02 Table B-3)Figure 4-4 Summary of Fire Area Review[Based on FAQ 07-0054 Revision I and modified per the VCSNS process]Compliance with NFPA 805 Requirements Page 35 Compliance with NFPA 805 Requirements Page 35

-EM RC-11-0149 Section 4.0 Results of the Evaluation Process Attachment C contains the results of the Fire Area Transition review (NEI 04-02 Table B-3). On a fire area basis, Attachment C summarizes compliance with Chapter 4 of NFPA 805. Attachment C also contains an overview of accomplishment of each of the performance goals.NEI 04-02 Table B-3 (Attachment C) includes the following summary level information for each fire area: " Regulatory Basis -NFPA 805 post-transition regulatory bases (4.2.3/4.2.4)." Performance Goal Summary -An overview of the method of accomplishment of each of the performance criteria in NFPA 805 Section 1.5 is provided." Reference Documents

-Specific references to NSCA Documents." DROIDs -Specific Deterministic Requirement Open Item Descriptions of NFPA 805 Section 4.2.3. References to disposition each DROID has been provided.Resolutions to NSCA DROIDs which been resolved by a deterministic -based methods have been summarized in Attachment C, and documented in VCSNS Technical Report TR08620-312 and include the following information to meet the deterministic requirements:

o Modifications

-Attachment S contains a list of required modifications for the fire area or fire zone in support of resolving compliance issues.o FPEEE -Specific references to FPEEE that rely on determinations of"adequate for the hazard" that will remain part of the post-transition licensing basis. A brief description of the condition and the basis for acceptability is provided.o Licensing Actions -Specific references to prior approved and credited exemption requests, deviations, and/or safety evaluations that will remain part of the post-transition licensing basis. A reference to those credited licensing actions are found in the FPEEE, with a brief description of the condition and the basis for acceptability of the Licensing Action found in Attachment K.Resolutions to NSCA DROIDs which have not been resolved by a deterministic-based methods, an assessment of viability to resolve using a performance based solution (VFDR) was completed.

When used, this approach to resolution has also been documented in Attachment C, and in VCSNS Technical Report TR08620-312.

These actions may include: o Performance-Based Evaluation:

Fire Modeling (VFDR) -A summary of the results of the Performance-Based Fire Modeling developed to disposition a VFDR, developed in accordance with NFPA 805, Section 4.2.4.1.o Performance-Based Evaluation:

Fire Risk Evaluations (VFDR) -A summary of the results of the Fire Risk Evaluations developed to disposition a VFDR, developed in accordance with NFPA 805, Section 4.2.4.2.Compliance with NFPA 805 Requirements Page 36 RC-11-0149 Section 4.0 o Recovery Actions -A summary of any required Recovery Actions necessary to disposition a DROID. References to the feasibility analysis are described in Section 4.2.1.3 of the Transition Report. See also Attachment G.Required FP Systems/ Features -Specific FP systems and features necessary to support the results of the analysis (see Section 4.8.2 of the Transition Report).4.3 Non-Power Operational Modes 4.3.1 Overview of Evaluation Process VCSNS implemented the process outlined in NEI 04-02 and FAQ 07-0040, Non-Power Operations Clarification.

The goal (as depicted in Figure 4-5) is to ensure that contingency plans are established when the plant is in a NPO mode where the risk is intrinsically high. During low risk periods, normal risk management controls and fire prevention/protection processes and procedures will be utilized.The process to demonstrate that the nuclear safety performance criteria are met during NPO modes involved the following steps: " Reviewed the existing Outage Management Processes." Identified Equipment/Cables necessary for each Mode of Station Operation:

o Reviewed plant systems and key components to define the success paths that support each of the defense-in-depth Key Safety Functions (KSFs), and o Identified cables required for the selected components and determined their routing." Performed Fire Area Assessments to define redundant functions that may be affected by a postulated fire in a fire area (pinch-points)." Developed strategies to manage pinch-points associated with fire-induced vulnerabilities during NPO modes.The process is depicted in Figures 4-5 and 4-6. The results are presented in Section 4.3.2 of the Transition Report.Compliance with NFPA 805 Requirements Page 37 Compliance with NFPA 805 Requirements Page 37 RC-11-0149 Section 4.0 RC-11-0149 Section 4.0 Figure 4-5 Review POSs, KSFs, Equipment, and Cables, and Identify Pinch Points Compliance with NFPA 805 Requirements Page 38 Compliance with NFPA 805 Requirements Page 38 Aýo RC-11-0149 Section 4.0 Higher Risk Evolution as Defined by Plant Specific Outage Risk Criteria for example 1) Time to Boil 2) Reactor Coolant System and Fuel Pool Inventory 3) Decay Heat Removal Implement Contingency Plan for Specific KSF Figure 4-6 Manage Pinch Points 4.3.2 Results of the Evaluation Process The non-power modes transition review was conducted in accordance with the methodology described in NFPA 805 Section 2.4.2 and Appendix B.2, NEI 04-02 Section 4.3.3 and Appendix F, and FAQ 07-0040. The results of the review are documented in VCSNS Technical Report TR07800-008, "Non-Power Operation Modes Transition Review and Table F-i." Compliance with NFPA 805 Requirements Page 39 A

  • RC-11-0149 Section 4.0 The Plant Operating States considered for equipment and cable selection are defined in the technical report. The report uses nuclear safety performance criteria for non-power modes of operation to establish Key Safety Functions (KSF). It then identifies a range of equipment that is available to satisfy the KSFs in support of the nuclear safety objectives.

The active components identified in the success path logic diagrams are consolidated into a single NPO equipment list. Interested stakeholders, including representatives from Engineering, Operations, Outage Management and PRA, formed an expert panel to review the KSFs, Success Paths and preliminary equipment spreadsheets.

The NPO integration methodology was validated and success paths and associated components were added or screened as necessary to assure a comprehensive NPO equipment list was produced.For the functional states identified, cable selection and circuit analysis was performed.

Circuit Analysis worksheets and drawing markups were developed, reviewed and verified, and the results were entered into the analytical models and databases.

Cable/Conduit routing was defined, as necessary, and updated into the VCSNS Cable Management System.An analysis was performed using compliance assessment software for each Fire Area using the "baseline" equipment set to determine whether each KSF could be met using the specified success paths. Each Fire Area was then analyzed to identify potential"Pinch Points" where fire induced failures could potentially prevent the achievement of a KSF. Approximately 50 "pinch points" were identified.

Compliance strategies that meet the deterministic requirements of NFPA 805 were developed.

If no deterministic compliance strategy could be established without the need for a plant modification, contingency actions were established.

These contingency actions include identifying alternate methods of recovering the affected equipment and changes to processes and procedures that will minimize the potential for the evolution of fires in areas of the plant critical to the operation of credited NPO equipment.

The compliance strategies and contingency actions used to recover the affected components were presented to a panel of operations and outage management personnel for concurrence.

The compliance assessment provides an accounting of recovery actions that would be required to be credited following implementation of the changes to outage processes, procedures or system design identified in the report.Additional changes to plant processes and/or procedures will be made to incorporate the contingency actions.See Attachment D for details. Based on incorporation of the recommendations from the technical report into appropriate plant procedures, during the implementation phase of the NFPA 805 transition, the performance goals for Non-Power Operations are fulfilled and the requirements of NFPA 805 are met.4.4 Radioactive Release Performance Criteria 4.4.1 Overview of Evaluation Process The review of the fire protection program against NFPA 805, Section 1.5.2 and FAQ 09-0056 for fire event related radioactive release was performed using the methodology Compliance with NFPA 805 Requirements Page 40 Compliance with NFPA 805 Requirements Page 40 A-'a°RC-11-0149 Section 4.0 documented in VCSNS Technical Report TR07800-006, "NFPA 805 Radioactive Release Report." The methodology is summarized as follows: " Reviewed fire pre-plans, fire brigade training materials, and plant procedures to identify fire protection program elements (e.g., systems / components

/procedural control actions / flow paths, etc.) that are being credited to meet the radioactive release goals, objectives, and performance criteria during all plant operating modes, including full power and non-power conditions." Reviewed engineering controls to ensure containment of gaseous and liquid effluents (e.g., smoke and fire fighting agents). This review included all plant operating modes (including full power and non-power conditions).

Otherwise, provided a bounding analysis, quantitative analysis, or other analysis that demonstrates that the limitations for instantaneous release of radioactive effluents specified in the unit's Technical Specifications are met.4.4.2 Results of the Evaluation Process The radioactive release review determined the fire protection program will be compliant with the requirements of NFPA 805 and the guidance in NEI 04-02 and RG 1.205 upon completion of the open items identified in Attachment E. See Attachment S, Table S-2, for the corresponding implementation items.The site specific review of the direct effects of fire suppression activities on radioactive release is summarized in Attachment E.The review determined that radiation release to any unrestricted area due to the direct effects of fire suppression activities (but not involving fuel damage) would be as low as reasonably achievable and would not exceed applicable 10 CFR, Part 20 and Part 50 limits.The main strategy for complying with the radioactive release requirements in NPFA 805 and the guidance in NEI 04-02 and RG 1.205 is ensuring that all buildings or areas containing radioactive hazards or the potential for an uncontrolled release during a fire have adequate strategies to minimize the uncontrolled release of radioactive material during fire fighting activities.

This includes the revision or creation of documentation such as fire pre-plans, fire brigade training materials, and fire emergency procedures.

This documentation is then managed through responsibilities defined in the Station Fire Protection Program, and implemented primarily through the Station Training Organization to ensure that NFPA 805 Radioactive Release objectives will continue to be met in the future.4.5 Fire PRA and Performance-Based Approaches RI-PB evaluations are an integral element of an NFPA 805 fire protection program. Key parts of RI-PB evaluations include: " A Fire PRA (discussed in Section 4.5.1 and Attachments U, V, and W of the Transition Report)." NFPA 805 Performance-Based Approaches (discussed in Section 4.5.2 of the Transition Report).Compliance with NFPA 805 Requirements Page 41 Compliance with NFPA 805 Requirements Page 41 A -6 RC-11-0149 Section 4.0 4.5.1 Fire PRA Development and Assessment In accordance with the guidance in RG 1.205, a Fire PRA model was developed for VCSNS in compliance with the requirements of Part 4 "Internal Fires at Power Probabilistic Risk Assessment Requirements," of the ASME and ANS combined PRA Standard, ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Application," (hereafter referred to as Fire PRA Standard).

VCSNS conducted peer reviews by independent industry analysts in accordance with RG 1.200 prior to a risk-informed submittal.

The resulting fire risk assessment model is used as the analytical tool to perform Fire Risk Evaluations during the transition process.Section 4.5.1.1 of the Transition Report describes the Internal Events PRA model.Section 4.5.1.2 describes the Fire PRA model. Section 4.5.1.3 describes the results and resolution of the peer reviews of the Fire PRA, and Section 4.5.1.4 describes insights gained from the Fire PRA.4.5.1.1 Internal Events PRA The VCSNS base internal events PRA Revision 6a was the starting point for the Fire PRA.The internal events PRA was modified to capture the effects of fire both as an initiator of an event and the subsequent potential failure modes for affected circuits or individual targets.The VCSNS internal events PRA had a peer review performed in August 2002 in accordance with guidance in NEI-00-02, Industry PRA Peer Review Process. All A & B level Findings and Observations from the WOG Internal Events PRA Peer Review have been addressed.

Although all C & D level findings have not been incorporated, all of the items that had the potential to significantly impact model results have been resolved.Following completion of sufficient work to address the Peer Review comments, a 2005 gap assessment of the VCSNS Internal Events PRA was performed to determine the scope of work required to ensure the VCSNS Internal Events PRA meets Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 1. The results of this review indicated that VCSNS had resolved most of the issues identified in the original peer review, but the review identified some F&Os that needed additional work as well as several new issues. Additionally (in this 2005 review) the VCSNS PRA was found to meet CC-Il or better for 211 of the 271 SRs from the ASME PRA Standard, but 45 of the elements were found to either not meet the requirement or to meet the requirements at a CC-I level. Following work at VCSNS to address the findings and to increase the capability category ratings of the elements that needed an upgrade to allow use of the model in risk informed applications, a focused review was performed as required by the ASME RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications" (and 2007 addenda ASME RA-Sc-2007, Appendix A). The conclusion of this 2007 focused review is that the model is of sufficient quality for use as a basis in developing a Fire PRA Model.The results are summarized in Attachment U.Compliance with NFPA 805 Requirements Page 42 Aý ° RC-11-0149 Section 4.0 4.5.1.2 Fire PRA The Fire PRA was conducted in accordance with the requirements of the combined PRA Standard, "ASME/ANS RA-Sa-2009," developed by the American Society of Mechanical Engineers (ASME) and the American Nuclear Society (ANS). The Fire PRA was performed using state-of-the-art PRA methodologies, including those developed jointly by the NRC's Office of Nuclear Regulatory Research (RES) and the Electric Power Research Institute (EPRI) that are described in NUREG/CR-6850/EPRI-101 1989 and Fire PRA related FAQs.The Internal Events PRA CDF and LERF models were used as the basis for the plant response model for fire, with modifications to account for fire-induced failures.

FRANX and Computer Aided Fault Tree Analysis (CAFTA) are the two primary software tools used for development of the Fire PRA model. A Fire PRA model is developed and quantified to obtain fire risk results. CDF and LERF models are quantified for each fire scenario.

In the initial quantification, CCDP is calculated for each fire scenario.

To quantify CCDPs, fire scenario frequencies are set to 1.0. To quantify overall CDF, fire scenario frequencies determined as part of the Fire PRA process are combined with the appropriate CCDPs for each fire scenario.

Details of the Fire PRA development are documented in "Fire PRA Plant Response Model Report," Attachment 4 to VCSNS Design Calculation DC00340-001, "Fire PRA Plant Final Report." Fire Model Utilization in the Application Fire modeling was performed as part of the Fire PRA development (NFPA 805 Section 4.2.4.1).

RG 1.205, Regulatory Position 4.2 and Section 5.1.2 of NEI 04-02, provide guidance to identify fire models that are acceptable to the NRC for plants implementing a risk-informed, performance-based licensing basis.The following fire models were used: " NUREG-1805, Fire Dynamic Tools (FDTs)" Consolidated Fire and Smoke Transport Model (CFAST)The acceptability of the use of these fire models is included in Attachment J.4.5.1.3 Results of Fire PRA Peer Review The VCSNS Fire PRA (VCS FRANX Model -FOR REPORT 1-21 .zip) was peer reviewed against the requirements of ASME/ANS RA-Sa-2009, Part 4. A peer review was conducted during the period of August 16, 2010 through August 20, 2010, and a follow-on peer review was conducted the week of February 21, 2011.All the finding F&Os were resolved and the resolutions incorporated into the Fire PRA model. The disposition of these F&Os and justification for the Fire PRA meeting Capability Category II of the combined ASME/ANS Fire PRA Standard (ASME/ANS RA-Sa-2009) is provided in Attachment V.4.5.1.4 Risk Insights Risk insights were documented as part of the development of the Fire PRA. The total plant fire CDF/LERF was derived using the NUREG/CR-6850 methodology for Fire PRA development and is useful in identifying the areas of the plant where fire risk is greatest.Compliance with NFPA 805 Requirements Page 43 A -'a RC-11-0149 Section 4.0 A review of the fire initiating events that collectively represent 95% of the calculated fire risk is included as Attachment W.4.5.2 Performance-Based Approaches NFPA 805 outlines the approaches for performing performance-based analyses.

As specified in Section 4.2.4 of the Transition Report, there are generally two types of analyses performed for the performance-based approach: " Fire Modeling (NFPA 805 Section 4.2.4.1)." Fire Risk Evaluation (NFPA 805 Section 4.2.4.2).4.5.2.1 Fire Modeling Approach Overview of Evaluation Process Fire Modeling Evaluations were completed as part of the VCSNS NFPA 805 transition.

These Fire Modeling Evaluations were developed using the process described in VCSNS design calculation, "Fire Modeling:

Generic Methodology." This methodology is based upon the requirements of NFPA 805, industry guidance in NEI 04-02, and RG 1.205.NFPA 805 Section 4.2.4.1 identifies the specific use of fire modeling as a performance-based method. The Fire Modeling Evaluation process consists of the following steps: " Step 1 -Identified the targets." Step 2 -Established damage thresholds." Step 3 -Determined limiting condition(s)." Step 4 -Established fire scenarios (Maximum Expected and Limiting)." Step 5 -Determined protection of required nuclear safety success path(s)." Step 6 -Provided operations guidance, as necessary.

The acceptance criteria for the Fire Modeling Evaluation consist of two parts." Target Damage -The fire modeling analysis defines and evaluates a postulated scenario involving the Maximum Expected Fire Scenario (MEFS). If target set damage does not occur then first acceptance criterion is met." MEFS<<LFS

-The performance of fire modeling involves a degree of uncertainty.

This uncertainty is addressed indirectly by the determination of the Limiting Fire Scenario (LFS). A comparison of MEFS and LFS is used to determine if a sufficient fire modeling margin exists. If sufficient fire modeling margin exists, then the fire modeling approach is acceptable.

A quantitative risk assessment does not have to be performed since qualitatively the conclusion can be made that the VFDR has a minimal impact on risk. (MEFS does not generate damage, and MEFS -LFS margin is sufficiently large to address uncertainties in modeling.)

Fire Model Utilization in the Application RG 1.205, Regulatory Position 4.2 and Section 5.1.2 of NEI 04-02, provide guidance on documenting the fire models used, and justifying that these fire models and methods are acceptable for use in performance-based analyses when performed by qualified Compliance with NFPA 805 Requirements Page 44 RC-11-0149 Section 4.0 users, have been verified and validated, and are used within their limitations and with the rigor required by the nature and scope of the analyses.

The following fire models were used: " NUREG-1805, Fire Dynamic Tools (FDTs)" Consolidated Fire and Smoke Transport Model (CFAST)The acceptability of the use of these fire models is included in Attachment J.Note: At VCSNS, the use of the fire modeling option (see NFPA 805, Section 4.2.4.1)to disposition potential variances from deterministic requirements follows a pre-defined process documented in the NFPA 805 project instructions or SCE&G design guides. The objective of these documents is to provide the framework for the use of fire modeling both during the NFPA 805 transition and in the future while the plant operates under NFPA 805 licensing basis. Consistent with the fire modeling requirements in NFPA 805, these documents allow for the use of fire models that are verified and validated within a range of applications.

Consequently, the fire models available for use at VCSNS when operating under NFPA 805 are not limited to the ones selected for supporting the transition.

Fire models that are verified and validated (e.g., FDS, CFAST, FDTs, etc.) and are exercised within the corresponding application range may be used in the future following the process outlined in the project instructions and in accordance with the requirements of NFPA 805.Results of Evaluation Process Disposition of VFDRs The VCSNS NFPA 805 transition project NSCA shutdown analysis has identified a number of VFDRs to NFPA 805 Section 4.2.3. A small number of these VFDRs were dispositioned using Performance-Based fire modeling.Each VFDR dispositioned using a Fire Model Evaluation was assessed against the Fire Model Evaluation acceptance criteria described NFPA 805, Section 2.4.1. The results of are summarized in the detailed fire modeling calculations for each analyzed fire area (Design Calculation series DC0780B-XXX).

4.5.2.2 Fire Risk Approach Overview of Evaluation Process The Fire Risk Evaluations were completed as part of the VCSNS NFPA 805 transition.

These Fire Risk Evaluations (FRE) were developed using the process described in the VCSNS Project Instruction, "PI 6.0 Fire Risk Evaluations." This methodology is based upon the requirements of NFPA 805, industry guidance in NEI 04-02, and RG 1.205.These are summarized in Table 4-5.Compliance with NFPA 805 Requirements Page 45 Compliance with NFPA 805 Requirements Page 45 R C 0149 Section 4.0 Table 4-5 Fire Risk Evaluation Guidance Summary Table Document Section(s)

Topic NFPA 805 -2001 2.2(h), 4.2.4, A.2.2(h), A.2.4.4, D.5 Change Evaluation (2.2(h), 2.2.9, 2.4.4 A.2.2(h), A.2.4.4, D.5)Risk of Recovery Actions (4.2.4)Use of Fire Risk Evaluation (4.2.4.2)NEI 04-02 Revision 2 4.4, 5.3, Appendix B, Appendix I, Change Evaluation, Change Evaluation Appendix J Forms (App. I), No specific discussion of Fire Risk Evaluation RG 1.205 Revision 1 C.2.2.4, C.2.4, C.3.2 Risk Evaluations (C.2.2.4)Recovery Actions (C.2.4)Note: Change evaluations will be used for post LAR. During transition, FREs were performed, which is the intent of the change evaluation.

During the transition to NFPA 805, selected VFDRs from Section 4.2.3 of NFPA 805 were dispositioned as a Fire Risk Evaluation per Section 4.2.4.2 of NFPA 805.If the Fire Risk Evaluation meets the acceptance criteria of RG 1.174, this is confirmation that CDF, LERF, delta CDF, and delta LERF are sufficiently low and that the performance-based approach is acceptable per Section 4.2.4.2 of NFPA 805.The Fire Risk Evaluation process consists of the following steps (Figure 4-7 depicts the Fire Risk Evaluation process used during transition, which is generally based on FAQ 07-0054 Revision 1): Step 1 -Preparation for the Fire Risk Evaluation." Definition of the Variances from the Deterministic Requirements.

The definition of the VFDR includes a description of problem statement and the section of NFPA 805 that is not met, type of VFDR (e.g., separation issue or degraded fire protection system), and proposed evaluation per applicable NFPA 805 section." Preparatory Evaluation

-Fire Risk Evaluation Team Review. Using the information obtained during the development of the NEI 04-02 B-3 Table and the Fire PRA, a team review of the VFDR was performed.

Depending on the scope and complexity of the VFDR, the team may include the Safe shutdown/NSCA Engineer, the Fire Protection Engineer, and the Fire PRA Engineer.

The purpose and objective of this team review was to address the following:

o Review of the Fire PRA modeling treatment of VFDR o Ensure discrepancies in the model were captured and resolved Step 2 -Performed the Fire Risk Evaluation The Evaluator coordinated as necessary with the Safe Shutdown/NSCA Engineer, Fire Protection Engineer and Fire PRA Engineer to assess the VFDR using the Fire Risk Evaluation process to perform the following:

o Change in Risk Calculation with consideration for additional risk of recovery actions and required fire protection systems and features due to fire risk.o Fire area change in risk summary Compliance with NFPA 805 Requirements Page 46 A -6 RC-11-0149 Section 4.0 Step 3 -Reviewed the Acceptance Criteria The acceptance criteria for the Fire Risk Evaluation consist of two parts. One is quantitatively based and the other is qualitatively based. The quantitative figures of merit are ACDF and ALERF. The qualitative factors are defense-in-depth and safety margin.o Risk Acceptance Criteria.

The transition risk evaluation was measured quantitatively for acceptability using the ACDF and ALERF criteria from RG 1.174, as clarified in RG 1.205 Regulatory Position 2.2.4.o Defense-in-Depth.

A review of the impact of the change on defense-in-depth was performed, using the guidance from NEI 04-02.o Safety Margin Assessment.

A review of the impact of the change on safety margin was performed.

Compliance with NFPA 805 Requirements Page 47 Compliance with NFPA 805 Requirements Page 47 Aýo RC-11-0149 Section 4.0 RC-11-0149 Section 4.0 Prepare for Fire Risk Evaluation Perform Fire Risk Evaluation Review of Acceptance Criteria Figure 4-7 Fire Risk Evaluation Process (NFPA 805 Transition)

[Based on FAQ 07-0054 Revision 1]Compliance with NFPA 805 Requirements Page 48 Compliance with NFPA 805 Requirements Page 48 RC-11-0149 Section 4.0 Results of Evaluation Process Disposition of VFDRs The VCSNS NFPA 805 transition project activities associated with the NSCA has identified a number of variances from the deterministic requirements of NFPA 805 Section 4.2.3. Some of these VFDRs have been dispositioned using the fire risk evaluation process.Each variance dispositioned using a Fire Risk Evaluation was assessed against the Fire Risk Evaluation acceptance criteria of ACDF and ALERF; and maintenance of defense-in-depth and safety margin criteria from Section 5.3.5 of NEI 04-02 and RG 1.205. The results of these calculations are summarized in Attachment C.Following completion of transition activities and planned modifications and program changes, the plant will be compliant with 10 CFR 50.48(c).Risk Change Due to NFPA 805 Transition In accordance with the guidance in RG 1.205, Section C.2.2.4, Risk Evaluations, risk increases or decreases for each fire area using Fire Risk Evaluations and the overall plant should be provided.

Note that the risk increase due to the use of recovery actions was included in the risk change for transition for each fire area.RG 1.205 Section C.2.2.4.2 states in part"The total increase or decrease in risk associated with the implementation of NFPA 805 for the overall plant should be calculated by summing the risk increases and decreases for each fire area (including any risk increases resulting from previously approved recovery actions).

The total risk increase should be consistent with the acceptance guidelines in Regulatory Guide 1.174. Note that the acceptance guidelines of Regulatory Guide 1.174 may require the total CDF, LERF, or both, to evaluate changes where the risk impact exceeds specific guidelines.

If the additional risk associated with previously approved recovery actions is greater than the acceptance guidelines in Regulatory Guide 1.174, then the net change in total plant risk incurred by any proposed alternatives to the deterministic criteria in NFPA 805, Chapter 4 (other than the previously approved recovery actions), should be risk neutral or represent a risk decrease." The risk increases and decreases are provided in Attachment W.4.6 Monitoring Program 4.6.1 Overview of NFPA 805 Requirements and NEI 04-02 Guidance on the NFPA 805 Fire Protection System and Feature Monitoring Program Section 2.6 of NFPA 805 states: "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." Compliance with NFPA 805 Requirements Page 49 Compliance with NFPA 805 Requirements Page 49 RC-11-0149 Section 4.0 The intent of the monitoring review is to confirm the adequacy of the surveillance, inspection, testing, compensatory measures, and oversight processes to support the Monitoring requirement defined in NFPA 805. This process considers the following: " Assessing the adequacy of the scope of structure, systems and components within existing plant programs" Defining the performance criteria for the availability and reliability of the required structure, systems and components" The adequacy of the plant corrective action program in determining causes of equipment and programmatic failures and in minimizing their recurrence

4.6.2 Overview

of Post-Transition NFPA 805 Monitoring Program This section provides an overview of the post-transition NFPA 805 monitoring program process. The monitoring program will be implemented after the safety evaluation issuance as part of the fire protection program transition to NFPA 805. The monitoring program described in this section is currently based on FAQ 10-0059 Revision 1, which has not yet been issued a closure memo. VCSNS will implement a monitoring program consistent with the NRC approved version of FAQ 10-0059.The monitoring process is comprised of five phases." Phase 1 -Scoping" Phase 2 -Screening Using Risk Criteria" Phase 3 -Risk Target Value Determination" Phase 4 -Monitoring Implementation" Phase 5 -Periodic Assessment Phase 1 -Scoping In order to meet the NFPA 805 requirements for monitoring, the following categories of SSCs and programmatic elements will be reviewed during the implementation phase for inclusion in the NFPA 805 monitoring program: " Structures, Systems, and Components required to comply with NFPA 805, specifically:

o Fire protection systems and features required by the Nuclear Safety Capability Assessment o Fire protection systems and features modeled in the Fire PRA o Fire protection systems and features required by Chapter 3 of NFPA 805 o Nuclear Safety Capability Assessment equipment o Structures, systems and components relied upon to meet radioactive release criteria" Fire Protection Programmatic Elements" Key Assumptions in Engineering Analyses (specifically analyses performed to demonstrate compliance with the nuclear safety and radioactive release performance criteria)Compliance with NFPA 805 Requirements Page 50 Compliance with NFPA 805 Requirements Page 50 RC-11-0149 Section 4.0 As a minimum the fire protection systems and features and SSCs required to meet the radioactive release criteria will be included in the existing inspection and test programs and in the system/program health program. In addition passive features (rated barriers, ERFBS), and other components (e.g. drains, curbs) that are relied upon to demonstrate compliance with Chapter 4 of NFPA 805 will also be included in the inspection and test programs, including system/program health reporting.

The existing programs are adequate for routine monitoring of these SSCs. SSCs that are not addressed in the existing programs will be added.Phase 2 -Screening Using Risk Criteria Phase 2 of the process uses the risk significance criteria and screens the SSCs and programmatic elements to determine High Safety Significant SSCs and programmatic elements.

This may be accomplished at the component, programmatic element, and/or functional level. Since risk is evaluated at the analysis unit level (e.g. fire compartment, fire area, fire scenario, ignition source), criteria must be developed to determine those analysis units for which the SSCs are considered High Safety Significant.

The Fire PRA is the primary tool used to establish the risk significance criteria and performance bounding guidelines.

The screening thresholds used to determine risk significant analysis units are those that meet the following criteria: Risk Achievement Worth (RAW) of the monitored parameter

> 2.0 (AND) either Core Damage Frequency (CDF) x (RAW) > 1.OE-7 per year (OR)Large Early Release Frequency (LERF) x (RAW) > 1.OE-8 per year High Safety Significant (HSS) fire protection systems and features and nuclear safety capability equipment are those that meet or exceed the risk significant screening criteria.

The SSCs and programmatic elements for these HSS analysis units will be included in the additional monitoring program of NFPA 805.Low Safety Significant fire protection systems and features and nuclear safety capability equipment are those that do not meet the risk significant screening criteria and are monitored via existing programs/processes.

Additionally, the review may include other analysis units (and required FP/NSCA SSCs and programmatic elements) that are not risk significant (per the screening criteria) but are included based on plant specific history and/or operational considerations.

Documentation of the High Safety Significance fire protection systems and features and nuclear safety capability equipment will be contained in an engineering or Fire PRA controlled document.Phase 3 -Risk Target Value Determination Phase 3 consists of using the Fire PRA, or other processes as appropriate, to determine target values of reliability and availability for the High Safety Significant, FP/NSCA SSCs and programmatic elements established in Phase 2.Compliance with NFPA 805 Requirements Page 51 RC-11-0149 Section 4.0 Failure criteria are established by an expert panel or evaluation based on the required fire protection and nuclear safety capability SSCs and programmatic elements assumed level of performance in the supporting analyses.

Action levels are established for the SSCs at the component level, program level, or functionally through the use of the pseudo system (or functional grouping concept).

The actual action level is determined based on the number of component, program or functional failures within a sufficiently bounding time period (-2-3 operating cycles). Adverse trends and unacceptable levels of availability, reliability, and performance will be reviewed against established action levels.Documentation of the monitoring program failure criteria and action level targets will be contained in a controlled document.

The basis for the criteria and action levels will be a controlled Engineering or Fire PRA evaluation.

It is anticipated that the availability and reliability criterion for High Safety Significant Performance Monitoring Groups will use the guidance included in several industry documents tempered by site-specific operating experience, Fire PRA assumptions, and equipment types (and vendor data or valid design input when available).

Industry documents such as the EPRI Fire Protection Equipment Surveillance Optimization and Maintenance Guide 1006756, Final Report July 2003, NFPA codes, and/or the NRC Fire Protection Significance Determination Process in addition to site specific operating experience data may be used.Phase 4 -Monitoring Implementation Phase 4 is the implementation of the monitoring program, once the monitoring scope and criteria are established.

The corrective action process will be used to address performance of fire protection and nuclear safety SSCs that do not meet performance criteria.For High Safety Significant fire protection and nuclear safety SSCs that are monitored, unacceptable levels of availability, reliability, and performance will be reviewed against the established action levels. If an action level is triggered, a non-conformance report or similar station document will be initiated to identify the negative trend. A corrective action plan will then be developed using an existing station process. An effective plan should improve performance, and return the SSC to above the established action level.Phase 5 -Periodic Assessment A periodic assessment will be documented and scheduled (e.g., at a frequency of approximately every two to three operating cycles), including, where practical, industry operating experience, and may be part of a larger assessment.

The objectives of this assessment include: " Review Systems with performance criteria.o Confirm performance criteria still effectively monitor the functions of the system o Confirm performance criteria still monitor the effectiveness of the fire protection and nuclear safety capability assessment systems" Configuration Control Compliance with NFPA 805 Requirements Page 52 RC-11-0149 Section 4.0 o Confirm via review of supporting analyses (e.g. Recent revisions) to determine if new fire protection features, NSCA SSCs, programmatic elements and/ or other functions should be added to monitoring scope o Confirm via review of supporting analyses (e.g. Recent revisions) to determine if the performance criteria is no longer applicable Trends o Determine if there any trends in system performance that are not being addressed, based on the performance during the assessment period.4.7 Program Documentation, Configuration Control, and Quality Assurance 4.7.1 NFPA 805 Documentation Requirements (NFPA 805, Section 2.7.1)In accordance with the requirements and guidance in NFPA 805 Section 2.7.1 and NEI 04-02, VCSNS has documented analyses to support compliance with 10 CFR 50.48(c).

The analyses and calculations have been performed in accordance with SCE&G's processes for ensuring assumptions are clearly defined, that results are easily understood, that results are clearly and consistently described, and that sufficient detail is provided to allow future review of the entire analyses.Documentation associated with compliance with 10 CFR 50.48(c) will be maintained for the life of the plant and organized to facilitate review for accuracy and adequacy.The Fire Protection Program Design Basis Document described in Section 2.7.1.2 of NFPA 805 and necessary supporting documentation described in Section 2.7.1.3 of NFPA 805 have been developed as part of transition to 10 CFR 50.48(c).

This documentation will be issued for use as part of program implementation following receipt of the license amendment.

Appropriate cross references have been established to supporting documents as required by SCE&G.Figure 4-8 depicts the planned post-transition documentation and relationships.

Compliance with NFPA 805 Requirements Page 53 Compliance with NFPA 805 Requirements Page 53 Aýýo RC-11-0149 Section 4.0 NFPA 805 DOCUMENTS-----------------..- .-. ------. .-------------------------.PCCKS DATABASE[NSCA Equip Sn~d Data Cables Routing PRA Equip and Non-Power Data (NPO) Equip and Data------------------ -----------

NSCA ANALYSIS COMP & Cable FA Assessment MethodlResults MethodlResults MSO and OMA NSCA Logic Treatments Diagrams NSCA SUPPORTING INFO Manual Action T-H Calculations Feasibility Coordination/

Common Erc/ Plant DBDs Flooding [Support NSCA]Calculations Revised License Condition Revised UFSAR B-2 Table I B-3 Table FIRE SAFETY ANALYSIS (DBD)On a Fire Area Basis-Fire Area Description

' Fire Hazards Information

'Nuclear Safety Performance Criteria Compliance Summary (NEI 04-02 B-3 Table Results)* Non-Power Evaluation Results Summary'Radioactive Release Summary On a Generic Basis'B-1 Table Results'Radioactive Release'Monitoring Program' Fire Hazards Analysis Drawings'Area Risk Summary Insights Non-Power Mode NSCA Treatment Non-Power Operations Analysi FM Database Ignition Sources Area Targets/& Scenarios Hazard Inventory Equipment/]

W eighting Fire Modeling Weighting FHA SUPPORT DOCS FP Administrative FP Condition Controls / Coniton Proceures Monitoring Procýedures FP Systems Code Compliance FP Drawings Reviews FP System and Engineering Feature DBI~s Equivalency Evaluations Radioactive Fire Pre-Plans Release Review FP Systems and B-I Table Features Info Detailed Data........................ .................. ........ ....NFPA 805 FIRE RISK EVALUATIONS Fire Risk Evaluation Report Note: Bold text indicates new NFPA 805 documents Figure 4-8 NFPA 805 Planned Post-Transition Documents and Relationships Compliance with NFPA 805 Requirements Page 54 Compliance with NFPA 805 Requirements Page 54 eWý* RC-11-0149 Section 4.0 4.7.2 NFPA 805 Configuration Management (NFPA 805, Sections 2.2.9/2.7.2)

Program documentation established, revised, or utilized in support of compliance with 10 CFR 50.48(c) is subject to SCE&G configuration control processes that meet the requirements of Sections 2.2.9 and 2.7.2 of NFPA 805. This includes the appropriate procedures and configuration control processes for ensuring that changes impacting the fire protection program are reviewed for impact. The RI-PB post transition change process methodology is based upon the requirements of NFPA 805, and industry guidance in NEI 04-02, and RG 1.205. These requirements are summarized in Table 4-6.Table 4-6 Change Evaluation Guidance Summary Table Document Section(s)

Topic NFPA 805 2.2(h), 2.2.9, 2.4.4, A.2.2(h), A.2.4.4, D.5 Change Evaluation NEI 04-02 5.3, Appendix B, Appendix I, Appendix J Change Evaluation, Change Evaluation Forms (Appendix I)RG 1.205 C.2.2.4, C.3.1, C.3.2, C.4.3 Risk Evaluation, Standard License Condition, Change Evaluation Process, Fire PRA The Plant Change Evaluation Process consists of the following 4 steps and is depicted in Figure 4-9: " Defining/Screening the Change" Performing the Preliminary Risk Screening" Performing the Risk Evaluation" Evaluating the Acceptance Criteria Change Definition The Change Evaluation process begins by defining the change or altered condition to be examined and the baseline configuration as defined by the Design Basis and Licensing Basis (NFPA 805 Licensing Basis post-transition).

1. The baseline is defined as that plant condition or configuration that is consistent with the Design Basis and Licensing Basis (NFPA 805 Licensing Basis post-transition).
2. The changed or altered condition or configuration that is not consistent with the Design Basis and Licensing Basis is defined as the proposed alternative.

Preliminary Risk Review Once the definition of the change is established, a screening is then performed to identify and resolve minor changes to the fire protection program. This screening is consistent with fire protection regulatory review processes in place at nuclear plants under traditional licensing bases. This screening process is modeled after the NEI 02-03 process. This process will address most administrative changes (e.g., changes to the combustible control program, organizational changes, etc.).The characteristics of an acceptable screening process that meets the "assessment of the acceptability of risk" requirement of Section 2.4.4 of NFPA 805 are: Compliance with NFPA 805 Requirements Page 55

° RC-11-0149 Section 4.0" The quality of the screen is sufficient to ensure that potentially greater than minimal risk increases receive detailed risk assessments appropriate to the level of risk." The screening process must be documented and be available for inspection by the NRC." The screening process does not pose undue evaluation or maintenance burden.If any of the above is not met, proceed to the Risk Evaluation step.Risk Evaluation The screening is followed by engineering evaluations that may include fire modeling and risk assessment techniques.

The results of these evaluations are then compared to the acceptance criteria.

Changes that satisfy the acceptance criteria of NFPA 805 Section 2.4.4 and the license condition can be implemented within the framework provided by NFPA 805. Changes that do not satisfy the acceptance criteria cannot be implemented within this framework.

The acceptance criteria require that the resultant change in CDF and LERF be consistent with the license condition.

The acceptance criteria also include consideration of defense-in-depth and safety margin, which would typically be qualitative in nature.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. In certain circumstances, an initial evaluation in the development of the risk assessment could be a simplified analysis using bounding assumptions provided the use of such assumptions does not unnecessarily challenge the acceptance criteria discussed below.Acceptability Determination The Change Evaluations are assessed for acceptability using the ACDF (change in core damage frequency) and ALERF (change in large early release frequency) criteria from the license condition.

The proposed changes are also assessed to ensure they are consistent with the defense-in-depth philosophy and that sufficient safety margins were maintained.

Compliance with NFPA 805 Requirements Page 56 Compliance with NFPA 805 Requirements Page 56 RC-11-0149 Section 4.0 Defining the Change (5.3.2)License No wt hp3o License Amendment r Request NOT Required Preliminary Risk Screening (5.3.3)Risk Evaluation (5.3.4)PRA Capability Category Assessment Fire PR.A Capability Category Assessment Acceptance Criteria (5.3.5)No Figure 4-9 Plant Change Evaluation

[NEI 04-02 Figure 5-1]Note references in Figure refer to NEI 04-02 Sections Compliance with NFPA 805 Requirements Page 57 Compliance with NFPA 805 Requirements Page 57 A- " RC-11-0149 Section 4.0 The VCSNS Fire Protection Program configuration is defined by the program documentation.

To the greatest extent possible, the existing configuration control processes for modifications, calculations and analyses, and Fire Protection Program License Basis Reviews will be utilized to maintain configuration control of the Fire Protection program documents.

The configuration control procedures which govern the various VCSNS documents and databases that currently exist will be revised to reflect the new NFPA 805 licensing bases requirements.

Several NFPA 805 document types such as: NSCA Supporting Information, Non-Power Operations Treatment, etc., generally require new control processes to be developed since they are new documents and databases created as a result of the transition to NFPA 805. The new processes will be modeled after the existing processes for similar types of documents and databases.

System level design basis documents will be revised to reflect the NFPA 805 role that the system components now play.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. 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 controlled documents depending upon the configuration control process being used. Reviews that identify potential Fire Protection program impacts will be sent to qualified individuals (Fire Protection, Safe Shutdown/NSCA, Fire PRA as applicable) to ascertain the program impacts, if any. 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:

Utilize 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.This process follows the requirements in NFPA 805 and the guidance outlined in RG 1.174 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.

4.7.3 NFPA 805 Quality Requirements (NFPA 805, Section 2.7.3)Fire Protection Program Quality During the transition to 10 CFR 50.48(c), VCSNS and its contract support staff performed work in accordance with the quality requirements of Section 2.7.3 of NFPA 805.Upon receipt of the NFPA 805 Safety Evaluation, VCSNS will implement a revised Quality Assurance Program to ensure compliance with section 2.7.3 of NFPA 805 within Compliance with NFPA 805 Requirements Page 58

  • e RC-11-0149 Section 4.0 the prescribed implementation period. The revised Fire Protection Quality Assurance Program is based on Regulatory Position 1.7, "Quality Assurance," in RG 1.189, Rev. 2,"Fire Protection for Operating Nuclear Power Plants." Fire PRA Quality Configuration control of the Fire PRA model will be maintained by integrating the Fire PRA model into the existing processes described in VCSNS procedure NL-126. This is the same procedure used to ensure configuration control of the internal events PRA model. This process complies with Section 5 of the ASME Standard for PRA Quality and ensures that VCSNS maintains an as-built, as-operated PRA model of the plant.The process has been peer reviewed.

Quality assurance of the Fire PRA is assured via the same processes applied to the internal events model.This process follows the guidance outlined in RG 1.174 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.

Although the entire scope of the formal 10 CFR 50 Appendix B program is not applied to the PRA models or processes in general, often parts of the program are applied. For instance, the procedure which addresses independent review of calculations for 10 CFR 50 Appendix B is applied to the PRA model calculations, as well.With respect to Quality Assurance Program requirements for independent reviews of calculations and evaluations, those existing requirements for Fire Protection Program documents will remain unchanged.

Quality: UncertaintylSensitivity As recommended by NUREG/CR-6850, the sources of uncertainty in the Fire PRA were identified and specific parameters were analyzed for sensitivity in support of the NFPA 805 FRE process.Specifically with regard to uncertainty, uncertainties associated with PRA assumptions and plant modifications associated with the project have been evaluated with sensitivity studies. These results are contained in VSCNS Design Calculation DC00340-001, Attachment 13, "FPRA Sensitivity and Uncertainty Report." In addition, sensitivity to uncertainty associated with fire initiating event frequencies is discussed in VCSNS Technical Report TR07800-007, "Fire PRA Ignition Frequency Analysis." Uncertainties associated with Fire Scenario Selection are associated in the notebooks for individual scenarios.

Quality: Conservatism While the removal of conservatism inherent in the Fire PRA is a long-term goal, the Fire PRA results were deemed sufficient for evaluating the risk associated with this application.

While VCSNS continues to strive toward a more "realistic" estimate of fire risk, use of mean values continues to be the best estimate of fire risk. During FRE process, the uncertainty and sensitivity associated with specific Fire PRA parameters were considerations in the evaluation of the change in risk relative to the applicable acceptance thresholds.

Compliance with NFPA 805 Requirements Page 59

-VE ° RC-11-0149 Section 4.0 Specific Quality Attributes (NFPA 805, Section 2.7.3)Review (NFPA 805, Section 2.7.3.1)Analyses, calculations, and evaluations performed in support of compliance with 10 CFR 50.48(c) were performed in accordance with VCSNS procedures that require independent review.Verification and Validation (NFPA 805, Section 2.7.3.2)Models and numerical methods used in support of compliance with 10 CFR 50.48(c)were verified and validated as required by Section 2.7.3.2 of NFPA 805.Limitations of Use (NFPA 805, Section 2.7.3.3)Engineering methods and numerical models used in support of compliance with 10 CFR 50.48(c) were and are used with the same limitations and assumptions supported by the V&V for the methods as required by Section 2.7.3.3 of NFPA 805.Qualification of Users (NFPA 805, Section 2.7.3.4)Cognizant personnel who use and apply engineering analysis and numerical methods in support of compliance with 10 CFR 50.48(c) was competent and experienced as required by Section 2.7.3.4 of NFPA 805. This requirement will continue to be met by adherence to SCE&G procedures and project management of contractor support staff.For personnel performing fire modeling or Fire PRA development and evaluation, VCSNS and contract personnel developed and maintained project instructions to be used by individuals assigned various tasks, to ensure consistency of the engineering and PRA products.

These instructions were developed by personnel with intimate knowledge and experience in the task subject matter. Task specific instructions were developed to identify and document required training and mentoring to ensure individuals are appropriately qualified per the requirements of NFPA 805 Section 2.7.3.4 to perform assigned work.Uncertainty Analysis (NFPA 805, Section 2.7.3.5)The impact of important uncertainties on the Fire PRA results was established using extensive, well formulated sensitivity studies to provide reasonable assurance that the performance criteria have been met as outlined in Section A of 2.7.3.5 of NFPA 805.4.8 Summary of Results 4.8.1 Results of the Fire Area Review A summary of the NFPA 805 compliance basis for each fire area is provided in Table 4-7. The table provides the following information from the NEI 04-02 Table B-3: " Fire Area: Fire Area Identifier."

Description:

Fire Area Description." NFPA 805 Compliance Basis: Post-transition NFPA 805 Chapter 4 compliance basis.The approach to performing the fire area reviews were done on a DROID by DROID basis. Deterministic based solutions were initially identified to disposition DROIDs.However, performance-based solutions were also selected on a case by case basis to Compliance with NFPA 805 Requirements Page 60 P 0 O-W _'"' ___ -4ý-9=MAMMY-RC-11-0149 Section 4.0 resolve DROIDs. If a given fire area has one performance based solution, then the overall area classification identified in Table 4-7 is performance-based.

If all solutions were deterministic, then the overall classification is deterministic.

Table 4-7 Fire Area NFPA 805 Compliance Basis Fire Area ABO1 CB01 CB02 CB03 CB04 CB05 CB06 CB07 CB08 CB10 CB12 CB14 CB15 CB 17 CB18 CB20 CB22 CB23 CWPHO1 CWPH02 CWPH03 Description AB General Area, All Elevations (ex WPAA)CB General Area 412, 425 West CB East Chase 400, 412 B Train CB West Chase 400, 412 CB East Lower Cable Spreading 425 CB East Chase 400, 412 B Train CB Relay Room CB Plant Computer Room CB General Area West/ West Chase 436, 448 CB East Chase 436 CB NE Chase 436 CB Security Computer Room CB Upper Cable Spreading Room CB Control Room/ Support Area CB East Chase 463 CB NE Chase 463 CB HVAC Room A CB HVAC Room B CWPH Electric Fire Pump Room CWPH Diesel Fire Pump Room CWPH Circ Water Pump Area (Outdoor)Underground Duct Bank (Conduits only)DG Diesel Generator A, All Elevations DG Diesel Generator B, All Elevations FH General Area, All Elevations 1B Battery Room X 412 IB Battery Room A 412 IB Battery Charger Room A 412 NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Compliance Basis NFPA 805 Section: 4.2.4 -Performance-Based Approach DS DG01 DG02 FHO1 IB01 IB02 IB03 NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.3 -Deterministic Approach Compliance with NFPA 805 Requirements Page 61 Compliance with NFPA 805 Requirements Page 61 RC-11-0149 Section 4.0 Table 4-7 Fire Area NFPA 805 Compliance Basis Fire Area IB04 IB05 IB06 IB07 IB08 IB09 IB10 IBll IB12 IB13 IB14 IB15 IB16 IB1 7 Description IB Battery Charger Room B 412 IB Battery Charger Room A/B 412 IB Battery Room B 412 IB Chilled Water Pump Rooms 412 IB HVAC Chiller Room C 412 IB HVAC Chiller Room B 412 lB Battery Room Ventilation A 423 IB SWBP Cooling Unit Room B IB Speed Switch Room B IB Speed/ XFR Switch Room "C" IB CREP Room A IB CREP Room B IB ESF SWGR Cooling Unit Room A IB ESF SWGR Cooling Unit Room B NFPA 805 Compliance Basis NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach IB18 IB Speed Switch Cooling Unit Room NFPA 805 Section: 4.2.3 -Deterministic Approach A IB19 IB Speed Switch Cooling Unit Room NFPA 805 Section: 4.2.3 -Deterministic Approach B IB20 IB 1 DA Switchgear Room NFPA 805 Section: 4.2.4 -Performance-Based Approach IB21 IB CRDM Switchgear Room NFPA 805 Section: 4.2.4 -Performance-Based Approach IB22 IB 1DB Switchgear (436)/ Ventilation NFPA 805 Section: 4.2.4 -Performance-Based Approach Room 423 IB23 lB A Chiller (412)/ SWBP Cooling NFPA 805 Section: 4.2.3 -Deterministic Approach (423)/ A Speed Switch Rooms (436)IB24 IB Reactor Protection Panel Room NFPA 805 Section: 4.2.3 -Deterministic Approach IB25 IB General Area 412, 436/ WPAA NFPA 805 Section: 4.2.4 -Performance-Based Approach 463 IB26 IB27 MH02 MH08 MH36 RB01 SWPH01 SWPH02 SWPH03 SWPH04 IB Electrical Chase 451 SE IB Diesel Generator B Cable Chase B train of MH02 Manhole Yard North Manhole Yard TBD RB General Area, All Elevations SWPH Elect Equip Room A SWPH Elect Equip Room C SWPH Elect Equip Room B SWPH Ventilation Duct Room NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.3 -Deterministic Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach NFPA 805 Section: 4.2.4 -Performance-Based Approach Compliance with NFPA 805 Requirements Page 62 Compliance with NFPA 805 Requirements Page 62 RC-11-0149 Section 4.0 Table 4-7 Fire Area NFPA 805 Compliance Basis Fire Area Description NFPA 805 Compliance Basis SWPH05 SWPH Service Water Pump Area NFPA 805 Section: 4.2.4 -Performance-Based Approach (436)/ Valve Pit Room (425)SWPH06 SWPH Cable Chase NFPA 805 Section: 4.2.3 -Deterministic Approach SWYD01 Electrical Switchyard (Outdoor)

NFPA 805 Section: 4.2.3 -Deterministic Approach TB01 TB General Area, All Elevations NFPA 805 Section: 4.2.4 -Performance-Based Approach 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 Tank NFPA 805 Section: 4.2.4 -Performance-Based Approach Area (Outdoor)YD02 Yard East of Plant (Outdoor)

NFPA 805 Section: 4.2.3 -Deterministic Approach YD03 Station Transformer Area NFPA 805 Section: 4.2.3 -Deterministic Approach 4.8.2 Required Fire Protection System/Feature Detection

/ suppression "required" for a given Fire Area is based on NFPA 805 Chapter 4 compliance.

The criteria are summarized below.o S -Separation Criteria:

Systems/Features required for Chapter 4 Separation Criteria in Section 4.2.3 o R -Risk Criteria:

Systems/Features required to meet the Risk Criteria for the Performance-Based Approach (Section 4.2.4)o E -EEEE/LA Criteria:

Systems/Features credited to support the acceptability of deviations found in Fire Protection Engineering Equivalency Evaluations

/NRC approved Licensing Action (i.e., Exemptions/Deviations/Safety Evaluations) (Section 2.2.7)o D -Defense-in-depth Criteria:

Systems/Features required to maintain adequate balance of Defense-in-Depth for a Performance-Based Approach (Section 4.2.4)The guidance used in the selection or "required" Fire Protection Systems/Features have been included in Table 4-8. A summary of the "required" fire suppression and detection systems have been included in Tables 4-9 and 4-10.The detailed listings of other required features (e.g. ERFBS, Fire Hose Stations, and Fire Hydrants) are available in controlled engineering documents.

The selection of these features has been made consistent with the guidance provided by Table 4-8, or other guidance provided by NFPA 805.The High Safety Significant (HSS) FP Systems/features are designated via the process defined in Section 4.6.2 of the Transition Report, as part of the Monitoring Program, and developed based on the NRC/NEI endorsed FAQ 10-0059 for defining these systems and features.Compliance with NFPA 805 Requirements Page 63 Compliance with NFPA 805 Requirements Page 63 RC-11-0149 Section 4.0 Attachment W contains the results of the Fire Risk Evaluations (including additional risk of recovery actions) and the change in risk on a fire area basis.Compliance with NFPA 805 Requirements Page 64 Compliance with NFPA 805 Requirements Page 64 A-* RC-11-0149 Section 4.0 Conditions for defining a Required Fire Protection System or feature shall assure that open issues or items where full, 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> separation (deterministic) cannot be achieved, and that the solution defines the appropriate system or feature to be required for alternative compliance approaches permitted by NFPA 805.Table 4-8 Required Fire Protection System Basis Fire Protection Deterministic (4.2.3) Performance-Based (4.2.4)Feature (NSCA) (Fire PRA/ Fire Model/ FRE)Fire Barriers 0 3 hr rated barrier for separation of 0 Controlled Fire Area envelop for a Fire Area/ Zone boundaries where redundant trains of equipment (4.2.3.2)

Performance-based fire model qualitative assessments are made* 3 hr enclosure of redundant System & 0 Credited Fire Rated barrier in the Base Fire regarding adequacy of separation by Components (4.2.3.3.a)

PRA or Fire Risk Evaluation (e.g. required to physical barriers and construction.

  • 1 hr enclosure of redundant System & meet the acceptance criteria of RG1.174 Components (4.2.3.3.c) guidelines) 0 1/2 hr radiant energy shield (4.2.3.4(b))

Fire Detection 9 Required to support area spatial (20 ft) Fire Detection system that is specifically

  • Fire Protection System or feature credited separation (4.2.3.3 (b)) credited in the Base Fire PRA or performance-in an Engineering Evaluation
  • Required to support area 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> separation based evaluation(e.g.

required to meet the (4.2.3.3 (c)) acceptance criteria of RG1.174 guidelines)

  • Credited for containment separation issue(4.2.3.4(c))

Fire Suppression 0 Required to support area spatial (20 ft) Fire Suppression System that is specifically

  • Fire Protection System or feature credited separation (4.2.3.3 (b)) credited in the Base Fire PRA or performance-in an Engineering Evaluation
  • Required to support area 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> separation based evaluation (e.g. required to meet the (4.2.3.3 (c)) acceptance criteria of RG1.174 guidelines) 0 Credited for containment separation issue(4.2.3.4(c))
  • Required by NFPA 805-2001 (e.g. Section 3.9.4)Hose Station
  • Hose stations in any area which are used to 9 Hose stations in any area which are used to
  • Fire Protection System or feature credited protect equipment in the NSCA model protect equipment in the Fire PRA model (e.g. in an Engineering Evaluation required to meet the acceptance criteria of RG1.174 guidelines)

Fire Hydrants

  • Fire Hydrants/

associated equipment used

  • Fire Hydrants/

associated equipment used as

  • Fire Protection System or feature credited as the primary means to support manual fire the primary means to support manual fire in an Engineering Evaluation fighting activities (water) fighting activities (water)Compliance with NFPA 805 Requirements Page 65 4rO4Wg:rM&G9 RC-11-0149 Section 4.0 Table 4-8 Required Fire Protection System Basis Fire Protection Deterministic (4.2.3) Performance-Based (4.2.4)Feature (NSCA) (Fire PRA/ Fire Model/ FRE)ERFBS & Credit being taken for separation of the Credit being taken in the Base Fire PRA or
  • Fire Protection System or feature credited affected equipment/

circuits by an ERFBS Performance-based evaluation for separation of in an Engineering Evaluation having a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (4.2.3.3(a))

or 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (4.2.3. the affected equipment/

circuits by a qualified 3(c)) fire resistance rating ERFBS (e.g. required to meet the acceptance 0 For containment separation issues, a 1/2 hr criteria of RG1.174)radiant energy shield (4.2.3.4(b))

or rated enclosure (4.2.3.4(b))

Notes: 1. The scope of Evaluations includes Fire Protection Engineering Equivalency Evaluations (2.2.7) and Defense in Depth discussions in Change Evaluations (2.4.4). In each case the FP system or feature becomes "required" when the system or feature is credited for supporting analysis conclusions (e.g. Defense in Depth, Supporting basis).2. Definition of a required FP system or feature implies compliance with the associated NFPA 805 Chapter 3 requirements.

3. A system or feature becomes "Required", when resolution is achieved by crediting one or more of these solutions on a fire zone or area basis.4. Refer to Attachment C for each fire area or zone, for identification of specific of "Required" FP systems and features.Compliance with NFPA 805 Requirements Page 66 Compliance with NFPA 805 Requirements Page 66 RC-11-0149 Section 4.0 Table 4-9 Summary of NFPA 805 Required Suppression Systems Required Fire Area / Fire Protected Area G S Reference Comments I ZoGeneral System Description Drawings NSCA Performance-Engineering References Based Evaluation Auxiliary Building AB.01 Fire Area AB01 Fire Area Assessment AB301.01 Gen Fl Area None None S Suppression AB374 "throughout fire area" TR0780E-001 AB01.02 RHR Pump Rm A None None S Suppression AB374 "throughout fire area" TR0780E-001 AB01.03 RHR Pump Rm B None None S Suppression AB374 "throughout fire area" TR0780E-001 AB301.04 General Corridor None None S Suppression Area/ Shield "throughout fire Slab, AB388 area" TR0780E-001 AB400 AB301.06 AB Charging None None S Suppression Pump Rm C, "throughout fire AB388 area" TR0780E-001 ABO1.08.01 Recirc Valve None None S Suppression Area North, "throughout fire AB397 area" TR0780E-001 ABO1.08.02 Recirc Valve None None S Suppression Area South, "throughout fire AB397 area" TR0780E-001 AB301.09 Charging Pump Preaction system, el 400', 1 MS-55-137 S E Partial Suppression HVAC Slab, Charging Pump HVAC Area for Room TR0780E-AB400 XVM-03428-FS 002, ECR50810 Compliance with NFPA 805 Requirements Page 67 Compliance with NFPA 805 Requirements Page 67 Aýý 8 RC-11-0149 Section 4.0 Table 4-9 Summary of NFPA 805 Required Suppression Systems Required Fire Area / Fire Protected Area G S Reference Comments I ZoGeneral System Description Drawings NSCA Performance-Engineering References Based Evaluation ABO1.21.01 Cable Tray Area Preaction system, el.463' South, 1 MS-55-137-4 S E Penetration Seals -South, AB463 Valve XVM-6940-FS AB423/CB423

-TR07870-002 ABO1.21.02 General Floor Preaction system, el.463' South, 1 MS-55-137-4 S E Penetration Seals -Area North, Valve XVM-6940-FS AB2145 -TR07870-AB463 002 Control Building CB01.01 North East Cable Wet-pipe system, el.412', Valve 1MS-55-085-25 S Chase, General XVM-4105-FS Floor Area (Below Ceiling)CB412 CB01.01 North East Cable Preaction system, Cable 1MS-55-085-10, S Chase, CB412 Chases, el.412', Valve XVM- 1 MS-55-085-11, 4065-FS 1 MS-55-085-14 CR01.02 Office Area West, Preaction system, Cable 1MS-55-085-10, S CB425 Spreading Areas, el.425', Valve 1 MS-55-085-11 XVM-4065-FS CB02 East Cable Preaction system, Cable 1 MS-55-085-10, S Chase, CB412 Chases, el.412', Valve XVM- 1 MS-55-085-11, 4065-FS 1 MS-55-085-14 CB04 Lower Cable Preaction system, Cable 1 MS-55-085-1 0, R Spreading Room, Spreading Areas, el.425', Valve 1MS-55-085-11 CB425 XVM-4065-FS CB05 East Cable Room Preaction Sprinkler System, 1MS-55-085-14 S& Pit Area CB 400', 412'400'&412" CB06 Relay Room, C02, Relay Rm, el.436', Room 1MS-55-040 R Compliance with NFPA 805 Requirements Page 68 A- 0 RC-11-0149 Section 4.0 Table 4-9 Summary of NFPA 805 Required Suppression Systems Required Fire Area / Fire Protected Area Reference Comments I General System Description Performance-Engineering References Zone Description Drawings NSCA Based Evaluation CB436 36-11 Sh.09 CB10 East Cable Preaction system, Cable 1 MS-55-085-10, S R E Penetration Seals -Chase, CB436 Chases, el.436', Valve XVM- 1MS-55-085-12, CB891 -TR07870-4065-FS 1 MS-55-085-14 002 CB12 Northeast Cable Preaction system, Cable 1 MS-55-085-10, S R E Penetration Seals -Chase, CB 436 Chases, el.436', Valve XVM- 1MS-55-085-12, CB1082, AB91/CB91 4065-FS 1 MS-55-085-14 -TR07870-002 CB15 Upper Cable Preaction system, Cable 1 MS-55-085-12 R E Penetration Seals -Spreading Room, Chases, el.463', Valve XVM- CB891, CB1082 -CB 448 4065-FS TR07870-002 CB18 East Cable Preaction system, Cable 1MS-55-085-10, R Chase, CB463 Chases, el.463', Valve XVM- 1 MS-55-085-13, 4065-FS 1 MS-55-085-14 CB20 North East Cable Preaction system, Cable 1 MS-55-085-10, E Penetration Seal -Chase, CB463 Chases, el.463', Valve XVM- 1 MS-55-085-13, CB423/AB423

-4065-FS 1 MS-55-085-14 TR07870-002 Circulating Water Pump House CWPH02 Diesel Fire Pump Wet-pipe system, Diesel fire 1MS-55-085-26 E Required by NFPA Room CWPH02 pump room, Valve XVG06817-805 Section 3.9.4.FS Room 36-02)Fuel Handing Building FH0I Fire Area FHOI Fire Area Assessment FHO1.03 FH Gen Fl Area None None S Suppression

& Tank Rooms "throughout fire 412 & 436 area" TR0780E-001 FHO1.04 FH Operating Fl None None S Suppression 436 "throughout fire Compliance with NFPA 805 Requirements Page 69 RC-11-0149 Section 4.0 Table 4-9 Summary of NFPA 805 Required Suppression Systems Required Fire Area / Fire Protected Area G S Reference Comments I ZoGeneral System Description Drawings NSCA Performance-Engineering References Based Evaluation area" TR0780E-001 I Intermediate Building IB07.01 Chilled Water Preaction system, 412' Chilled 1MS-55-137-5, S Separation between Pump Area Water Pump A, Valve XVM- 1 MS-55-137-6, Pumps, TR0780E-South, IB412 6935-FS 1MS-55-137-6A 001 IB07.02 Chilled Water Preaction system, 412' Chilled 1 MS-55-137-5, S Separation between Pump Area Water Pump B, Valve XVM- 1 MS-55-137-6, Pumps, TR0780E-North, IB412 6935-FS 1MS-55-137-6A 001 IB07.03 Chilled Water Preaction system, 412' Chilled 1MS-55-137-5, S Separation between Pump Area Water Pump C, Valve XVM- 1MS-55-137-6, Pumps, TR0780E-Central, IB412 6935-FS 1MS-55-137-6A 001 IB 25 Fire Area IB25 Fire Area Assessment IB25.01.01 SW Booster Preaction system, 412' and 436' 1 MS-55-137-5, S Separation between Pump A, (Below SW, Valve XVM-6935-FS 1MS-55-137-6, Pumps,TR0780E-Suspended 1 MS-55-137-6A 001 Barrier) West, IB 412 IB25.01.02 General Floor Preaction system, 412' and 436' 1MS-55-137-5, S R Separation between Area West, IB412 SW, Valve XVM-6935-FS 1MS-55-137-6, Pumps,TR0780E-1 MS-55-137-6A 001 IB25.01.03 General Floor Preaction system, 412' and 436' 1MS-55-137-5, S R Separation between Area East, IB412 SW, Valve XVM-6935-FS 1 MS-55-137-6, Pumps,TR0780E-1 MS-55-137-6A 001 IB25.01.04 Outside Turbine Preaction system, 412' and 436' 1MS-55-137-5, S EFW East Area, SW, Valve XVM-6935-FS 1 MS-55-137-6, IB 412 1 MS-55-137-6A Compliance with NFPA 805 Requirements Page 70 Compliance with NFPA 805 Requirements Page 70 RC-11-0149 Section 4.0 Table 4-9 Summary of NFPA 805 Required Suppression Systems Required Fire Area / Fire Protected Area Reference Comments I ZoeDsrpin General System Description DaigNSA Performance-Engineering References Zone Description Drawings NSCA BaeEvltin Rfrcs Based Evaluation 1B25.01.05 General Floor Preaction system, 412' and 436' 1MS-55-137-5, S R Separation between Area Central, SW, Valve XVM-6935-FS 1MS-55-137-6, Pumps,TR0780E-IB412 1 MS-55-137-6A 001 1B25.03.01 General Area None None S Suppression EPAA North, "throughout fire EPAA412 area" TR0780E-001 1B25.03.02 General Area None None S Suppression EPAA South, "throughout fire EPAA412 area" TR0780E-001 1B25.06.01 General MSIV None None S Suppression Area Room 36- "throughout the area" 02 TR0780E-001 1B25.06.02 General Area Preaction system, 436', Valve 1MS-55-137-5 E Penetration Seal -Room 36-02W XVM-6935-FS IB183 TR07870-002 1B25.07 General Area None None S Suppression South Room 36- "throughout the area" 02 TR0780E-001 Service Water Pump House SWPH05.01.01 Service Water None None S Separation

&Pump Discharge Suppression Valve Area, TR0780E-001 SWPH425 SWPH05.01.02 Service Water None None S Separation, Pump Discharge Detection

&Valve Area, Suppression SWPH425 TR0780E-001 Compliance with NFPA 805 Requirements Page 71 Compliance with NFPA 805 Requirements Page 71 RC-11-0149 Section 4.0 Table 4-9 Summary of NFPA 805 Required Suppression Systems Required Fire Area I Fire Protected Area Reference Comments I ZoeDsrpin General System Description DaigNCA Performance-Engineering References Zone Description Drawings NSCA PromneRfrne Based Evaluation SWPH05.01.03 Service Water None None S Separation, Pump Discharge Detection

&Valve Area, Suppression SWPH425 TR0780E-001 SWPH05.02.01 SWPH Pump Rm Preaction System, Service 1MS-55-137-1 S Separation, A South 436 Water Pumphouse 436' & 441', TR0780E-001 Valve XVM-6942-FS SWPH05.02.02 SWPH Pump Rm Preaction System, Service 1MS-55-137-1 S Separation, C Center 436 Water Pumphouse 436' & 441', TR0780E-001 Valve XVM-6942-FS SWPH05.02.03 SWPH Pump Rm Preaction System, Service 1MS-55-137-1 S Separation, B North 436 Water Pumphouse 436' & 441', TR0780E-001 Valve XVM-6942-FS Yard Areas YD02.01 Condensate None None S Suppression Storage Tank -"throughout fire South Side area" TR0780E-001 YD02.02 Condensate None None S Suppression Storage Tank -"throughout fire North Side area" TR0780E-001 Compliance with NFPA 805 Requirements Page 72 Compliance with NFPA 805 Requirements Page 72 RC-11-0149 Section 4.0 Table 4-10 Summary of NFPA 805 Required Detection Systems Fire Area /Required Fire Zone Protected Area Description NSCA Performance-Engineering Comments I References Based Evaluation Auxiliary Building AB.01 Fire Area Assessment ABO1.01.01 Auxiliary Bldg, el.374', Room 74-09S S Separation TR0780E-001 ABO1.01.02 Auxiliary Bldg, el.374', Room 74-09NE, -S Separation TR0780E-001 11, -12, -13, -14 ABO1.01.03 Auxiliary Bldg, el.374', Rooms 74-01, -07, -S Separation TR0780E-001 08, -09N, -09W and -18 ABO1.02 Auxiliary Bldg, el.374', Room 74-17 S Separation TR0780E-001 ABO1.03 Auxiliary Bldg, el.374', Room 74-16 S Separation TR0780E-001 ABO1.04 Auxiliary Bldg, el.388', Rooms 88-05, -05E, S R Separation TR0780E-001

-05W, -13, -13N, -13S, -13NE and -16 ABOI.04 Auxiliary Bldg, el.397', Room 97-01 S R Separation TR0780E-001 ABO1.04 Auxiliary Bldg, el.400', Rooms 00-01, 00- R Separation TR0780E-001 01E, and 00-01W ABO1.06 Auxiliary Bldg, el.388', Room 88-24 S Separation TR0780E-001 ABO1.08.01 Auxiliary Bldg, el.397', Room 97-02S S Separation TR0780E-001 ABO1.08.02 Auxiliary Bldg, el.397', Rooms 97-02, -02N S R Separation TR0780E-001 ABO1.09 Auxiliary Bldg, el.400', Room 00-02E S R Separation TR0780E-001 ABO1.10 Auxiliary Bldg, e1.412', Rooms 12-09, 12- R 28,26-02E, 26-02W ABO1.10 Auxiliary Bldg, el.412', Rooms 12-11, 12- R 11N ABO1.10 Auxiliary Bldg, el.426', Rooms 26-01, -02E, R Compliance with NFPA 805 Requirements Page 73 mwm--ý4rN%4%CM&G@

RC-11-0149 Section 4.0 Table 4-10 Summary of NFPA 805 Required Detection Systems Fire Area Required Fire Zone Protected Area Description NSCA Performance-Engineering Comments / References Based Evaluation

-02W ABO1.17 Auxiliary Bldg, el.412°, Rooms 12-02, -03A R ABO1.18.01 Auxiliary Bldg, el.436', Rooms 36-18, -17E R E Penetration Seals -AB 114/IB114(36-18)

-TR07870-002 ABO1.18.02 Auxiliary Bldg, el.436', Rooms 36-01, -03, -R E Penetration Seal -CB91/AB91 (AB36-31)31, -33 TR07870-002 ABO1 .18.02 Auxiliary Bldg, el.446' & 448', Rooms 46- R 01,48-01 AB01.18.02 Auxiliary Bldg, el.452', Rooms 52-01, -02 R ABO1.21.01 Auxiliary Bldg, el.463', Room 63-19 S R Separation TR0780E ABO1.21.02 Auxiliary Bldg, el.463', Rooms 63-04, -07, -S R Separation TR0780E 14, -16, and -17 ABO1.21.02 Auxiliary Bldg, el.463', Room 63-09 S R E Penetration Seal -AB423/CB423, AB 2145 (63-16) -TR07870-002 ABO1.29 Auxiliary Bldg, el.463', Room 63-01 R E Penetration Seal -AB2145 TR07870-002 ABO1.30 Auxiliary Bldg, el.485', Room 85-01, -02, S and -03 Control Building CB01.01 Control Bldg, el.412', Cable Chase, Rooms R 12-03, -11 CB01.01 Control Bldg, el.412', 12 -03 (above ceiling) S R CB01.02 Control Bldg, el.425', Rooms 25-01, -03, -S 04 CB02 Control Bldg, el.400' & 412', Cable Chase, S R Rooms00-01A & 12-04 Compliance with NFPA 805 Requirements Page 74 "W: " RC-11-0149 Section 4.0 Table 4-10 Summary of NFPA 805 Required Detection Systems Fire Area IRequired Protected Area Description Performance-Engineering Comments I References Fire Zone Prtce raDsrpinNSCA Based Evaluation CB04 Control Bldg, el.425', Room 25-02 R CB05 Control Bldg, el.400' & 412', Cable Chase, S R Rooms 00-01 & 12-04A CB06 Control Bldg, el.436', Room 36-11 R CB08.05 Control Bldg, el. 448', Room 48-01A E Penetration Seal -CB228 -TR07870-002 CB10 Control Bldg, el.436', Cable Chase, Room R E Penetration Seal -CB891 -TR07870-002 36-04 CB12 Control Bldg, el.436', Cable Chase, Room S R E Penetration Seal -CB1082 -TR07870-002 36-03 CB15 Control Bldg, el.448', Room 48-02 R E Penetration Seal -CB891, CB1 082 -TR07870-002 CB17.01 Control Bldg, el.463', Room 63-05 in MCB R CB17.01 Control Bldg, el.463', Rooms 63-05, -13 R CB17.02 Control Bldg, el.463', Rooms 63-06, -07, -E Penetration Seal -CB211 -TR07870-002 10, -11,-12 CB18 Control Bldg, el.463', Room 63-04 R CB20 Control Bldg, el.463', Room 63-03 R E Penetration Seal -CB423/AB423

-TR07870-002 CB22 Control Bldg, e1.482', Rooms 82-02, -03 E Penetration Seal -CB21 1- TR07870-002 CB23 Control Bldg, e1.482', Rooms 82-01, -04 E Penetration Seal -CB228- TR07870-002 Diesel Generator Building DG01.01 Diesel Generator Bldg, el.400', Room 00- E Penetration Seal UMS DG0001 -TR07870-002 01 Fuel Handing Building Compliance with NFPA 805 Requirements Page 75 RC-11-0149 Section 4.0 Table 4-10 Summary of NFPA 805 Required Detection Systems Fire Area / Required Protected Area Description Performance-Engineering Comments / References Fire Zone NSCA Promne niern Based Evaluation FHOI Fire Area FHOI FHO1.01 Fuel Handling Bldg, el.412', Room 12-01 Detection TR0780E-001; Penetration Seal UMS S E FH 1201 RBW/RB1201 NNWN -TR07870-002 FHO1.03 Fuel Handling Bldg, el.436', Rooms 36- Detection TR0780E-001; Penetration Seal UMS 01E, -01W S E RB3601NNWN/FH3601W

-TR07870-002 FHO1.03 Fuel Handling Bldg, el.443'-6", Room 443- S Detection TR0780E-001 01 FHO1.04 Fuel Handling Bldg, el.463', Rooms 63-01, Detection TR0780E-001; Penetration Seal UMS-01N, -01S S E RB6301NNWW/FH6301SSRBW

-TR07870-002 Intermediate Building 1B07.01 Intermediate Bldg, el.412', Room 12-13C S Separation TR0780E-001 1B07.02 Intermediate Bldg, el.412', Room 12-13B S Separation TR0780E-001 1B07.03 Intermediate Bldg, el.412°, Room 12-13A S Separation TR0780E-001 IB10 Intermediate Bldg, el.423', Room 23-02 R IB113 Intermediate Bldg, el.426', Room 26-01 R IB14 Intermediate Bldg, el.436', Room 36-03A R E Penetration Seals -IB183 -TR07870-002 IB17 Intermediate Bldg, el.451', Room 51-02 R IB20 Intermediate Bldg, el.463', Room 63-01 R 1B21.01 Intermediate Bldg, el.463', Room 63-02 R 1B21.02 Intermediate Bldg, el.463', Room 63-03 R 1B22.01 Intermediate Bldg, el.423', Room 23-01 R Compliance with NFPA 805 Requirements Page 76 Compliance with NFPA 805 Requirements Page 76 4 P 0 =_1 --- -ý-'3CZAMM-RC-11-0149 Section 4.0 Table 4-10 Summary of NFPA 805 Required Detection Systems Fire Area Required Fire Zone Protected Area Description NSCA Performance-Engineering Comments I References Based Evaluation 1B22,02 Intermediate Bldg, el.436', Room 36-01 R 1B23.02 Intermediate Bldg, el.426', Room 26-02 R IB24 Intermediate Bldg, el.436', Room 36-03B R IB25 Fire Area Assessment 1B25.01.01 Intermediate Bldg, el.412', Room 12-02W S Separation TR0780E-001 (SWBP A)1B25.01.02 Intermediate Bldg, el.412', Room 12-02 S R Separation TR0780E-001 1B25.01.02 Intermediate Bldg, el.412', Room 12-02W S R Separation TR0780E-001 1B25.01.03 Intermediate Bldg, el.412', Room 12-02 S R Separation TR0780E-001 1B25.01.03 Intermediate Bldg, el.412', Room 12-02E S R Separation TR0780E-001 1B25.01.03 Intermediate Bldg, el.423', Room 236-01 S R Separation TR0780E-001 1B25.01.04 Intermediate Bldg, el.412', Room 12-02E S 1B25.01.05 Intermediate Bldg, el.412', Room 12-02 S R Separation TR0780E-001 1B25.03.01 Intermediate Bldg, el.412', East Separation TR0780E-001; Penetration Seal -Penetration Area North, Room 12-01 S E UMS PAI1201W/Rb1201 EE -TR07870-002 1B25.03.02 Intermediate Bldg, el.412', East S Separation TR0780E-001 Penetration Area South, Room 12-01 1B25.04 Intermediate Bldg, el.412', West E Penetration Seal -UMS PAI1201E/RB1201WW

-Penetration Area South, Room 12-01 TR07870-002 1B25.05.01 Intermediate Bldg, el.436', East E Penetration Seal -UMS Penetration Area North, Room 36-01 RB3601 NSS/PA13601 RBW -TR07870-002 1B25.06.01 Intermediate Bldg, el.436', Room 36-02 S 1B25.06.02 Intermediate Bldg, el.436', Room 36-02 S R E Penetration Seal -1B183, AB114/1B114

-Compliance with NFPA 805 Requirements Page 77 A-6 RC-11-0149 Section 4.0 Table 4-10 Summary of NFPA 805 Required Detection Systems Fire rea /Required Fire Area Protected Area Description Performance-Engineering Comments I References Fire Zone NSCA Promne niern Based Evaluation TR07870-002 1B25.07 Intermediate Bldg, el.436', Room 36-02 S Suppression "throughout the area" TR0780E-001 11325.09 Intermediate Bldg, el.463', Room 63-03 E Penetration Seal -UMS PAA3601 E/RB3601WW

-6E TR07870-002 IB27 Intermediate Bldg, el.412', Room 12-09 S Separation TR0780E-003 Reactor Building RB01.01.01 Reactor Bldg, el.412', Rooms 12-01NNW, E Penetration Seals -UMS PAA1201E/RB1201WW 12-01NW,12-01W, and 12-01SW -TR0787E-01 RB01.01.02 Reactor Bldg, el.412', Rooms12-01S, 12- E Penetration Seals -UMS PAI1201W/RB1201EE

-01SE,12-01E, 12-NE, and 12-01NNE TR0787E-01 RB01.03.01 Reactor Bldg, el.436', Room 36-01 NNW E Penetration Seals -UMS RB360WNWN/FH3601W

-TR07870-002 RB01.03.02 Reactor Bldg, el.436', Rooms 36-01NNW, E Penetration Seals -UMS PAA3601 E/RB3601WW 36-01NW,36-01W, and 36-01SW -TR0787E-01 RB01.03.03 Reactor Bldg, el.436', Rooms 36-01SE, 36- E Penetration Seals -UMS 01S,36-01E, 36-01NE, 36-01NNE RB3601 NSS/PA13601 RBW -TR0787E-01 RB01.04.01 Reactor Bldg, el.463', Rooms 63-01NNW, E Penetration Seals -UMS RB6301WW/PAA6303E 63-01NNE -TR0787E-01 Turbine Building TB02 Turbine Bldg, el.412', Room 12-01 R E Penetration Seals -TB27 -TR07870-002 TB-03 Turbine Bldg, el.436', Room 36-01 R Yard0 YD02.01 Condensate Storage Tank -South Side S Detection, TR0780E-001 YDO2.02 Condensate Storage Tank -North Side S Detection, TRO780E-001 Compliance with NFPA 805 Requirements Page 78 RC-11-0149 Section 4.0 Table 4-10 Summary of NFPA 805 Required Detection Systems Fire Area IRequired Fire Are Protected Area Description Performance-Engineering Comments I References Based Evaluation Service Water Pump House SWPHI01 Service Water Pumphouse, el.425', Room R 25-05 SWPH03 Service Water Pumphouse, el.441', Room E Penetration Seals -SW341 -TR07870-002 41-01 SWPH04.01 Service Water Pumphouse, el.441', Room E Penetration Seals -SW341 -TR07870-002 41-01 A SWPH05.01.01 Service Water Pumphouse, el.425', Room 25-03 SWPH05.01.02 Service Water Pumphouse, el.425', Room 25-01 SWPH05.01.03 Service Water Pumphouse, el.425', Room 25-02 SWPH05.02.01 Service Water Pumphouse, el.436', Room 36-01 SWPH05.02.02 Service Water Pumphouse, el.436', Room 36-01 SWPH05.02.03 Service Water Pumphouse, el.436', Room 36-01 S S S S S S Separation TR0780E-001 Separation TR0780E-001 Separation TR0780E-001 Separation TR0780E-001 Separation TR0780E-001 Separation TR0780E-001 Compliance with NFPA 805 Requirements Page 79 Compliance with NFPA 805 Requirements Page 79 IEr o RC-11-0149 Section 4.0 4.8.3 Fire Risk Insights Fire PRA Overall Risk Insights Risk insights were documented as part of the development of the Fire PRA. The total plant fire CDF/LERF was derived using the NUREG/CR-6850 methodology for Fire PRA development and is useful in identifying the areas of the plant where fire risk is greatest.The risk insights generated were useful in identifying areas where specific contributors might be mitigated via modification.

A detailed description of significant risk sequences associated with the fire initiating events that collectively represent 95% (and individually any sequences above 1 % contribution) of the calculated fire risk for the plant was prepared for the purposes of gaining these insights and an understanding of the risk significance of MSO combinations.

These insights are provided in Attachment W, Tables W-1 and W-2.Risk Change Due to NFPA 805 Transition In accordance with the guidance in Regulatory Position 2.2.4.2 of RG 1.205 Revision 1: "The total increase or decrease in risk associated with the implementation of NFPA 805 for the overall plant should be calculated by summing the risk increases and decreases for each fire area (including any risk increases resulting from previously approved recovery actions).

The total risk increase should be consistent with the acceptance guidelines in Regulatory Guide 1.174. Note that the acceptance guidelines of Regulatory Guide 1.174 may require the total CDF, LERF, or both, to evaluate changes where the risk impact exceeds specific guidelines.

If the additional risk associated with previously approved recovery actions is greater than the acceptance guidelines in Regulatory Guide 1.174, then the net change in total plant risk incurred by any proposed alternatives to the deterministic criteria in NFPA 805, Chapter 4 (other than the previously approved recovery actions), should be risk-neutral or represent a risk decrease." Delta risk calculations are performed on the fire areas that have recovery actions, or for which the Nuclear Safety Capability Assessment (NSCA) identifies variances from deterministic requirements (VFDRs) to be addressed with a fire risk evaluation (FRE).The corresponding fire areas are as follows: ABO1, CB01, CB02, CB03, CB04, CB05, CB06, CB15, CB17, CB18, CB20, CB22, CB23, IB05, IB10, 1B12, 1B13, 1B14, IB15, 1B17, 1B20, 1B21, 1B22, 1B25, 1B27, RB01, SWPH03, SWPH04, SWPHO5, TBO1, TB02, and YD01.The total delta risks (fire-induced and from internal events) are found to be, with an acceptable safety margin, within the acceptable limits of Regulatory Guide 1.174, namely in Region III of Figure 3 and Figure 4 of that guide (i.e., delta CDF less than 1 E-06/yr and delta LERF less than 1 E-07/yr).

In addition, a qualitative analysis of DID supported by Fire PRA insights finds an adequate balance between the DID echelons, which do not require further improvements.

The fire-induced CDF and LERF at the plant level are approximately equal to 5.7E-05/yr and 2.7E-07/yr, respectively.

In addition, the internal-event (including internal flood)contributions to the CDF and LERF at the plant level are approximately 3.6E-06/yr and 1.3E-07/yr, respectively.

This results in a total baseline CDF and LERF approximately Compliance with NFPA 805 Requirements Page 80

-ZA o RC-11-0149 Section 4.0 equal to 6E-05/yr and 4E-07/yr.

These numbers credit the planned plant modifications (

References:

PRA Evaluations 11-4 and 11-13).At the plant level, the cumulative delta CDF and delta LERF, accounting for both VFDRs and recovery actions, are approximately equal to 4.6E-06/yr and 8.OE-09/yr, respectively.

These cumulative delta risks can further be broken down into their contributions from VFDRs and recovery actions. Namely, the cumulative delta CDF (delta LERF) from the VFDRs is approximately equal to 3.6E-06/yr (5.3E-09/yr), and the cumulative delta CDF (delta LERF) from the recovery actions is approximately equal to 9.6E-07/yr (2.7E-09/yr).

These delta risks are based solely on the scope of the fire initiating events. In addition, the delta CDF and delta LERF from internal events (including internal flood) between the post-transition and the pre-transition plant are equal to -8.8E-06/yr and -3.1E-09/yr, respectively (the decrease in risk is due to the modifications between the pre- and post-transition plants). This sums to a global delta CDF and delta LERF respectively equal to: -4.2E-06/yr and 4.9E-09/yr (

Reference:

PRA Evaluation 11-4).The cumulative delta risk values are in compliance with the numerical performance criteria of Regulatory Guide 1.174.4.8.4 Plant Modifications and Items to be Completed During the Implementation Phase Planned hardware modifications to comply with NFPA 805 are described in Attachment S, Table S-1.The VCS Fire PRA or Engineering teams did not identify any planned plant changes that would significantly impact the PRA model (see Section 4.0 of the Transition Report), beyond those identified and scheduled to be implemented as part of the transition to the 10 CFR 50.48(c) FPP, as set forth in the license condition.

The Fire PRA model represents the as-built, as-operated and maintained plant, including known proposed plant changes identified through ADD DATE, as it will be configured at the completion of the transition to NFPA 805. The Fire PRA model includes credit for the planned implementation of new improved RCP seal packages and an alternate seal injection (ASI) system, as well as modifications to existing operating procedures.

Following installation of new RCP seal packages and ASI, and the attendant installation details, additional refinements related to the new reactor coolant pump (RCP) seal packages and ASI modifications may need to be incorporated into the FPRA model. The same is true of the new procedures once they are finalized.

However, these changes are not expected to be significant and will likely result in additional risk improvement.

No other significant plant changes are outstanding with respect to their inclusion in the Fire PRA model. Additional modifications discussed in Attachment S, Table S-1, are also included in the FPRA model and their effect on the fire risk quantification results is included.

If significant plant changes are implemented but were not previously incorporated into the Fire PRA model, they will be screened, dispositioned and scheduled for incorporation into the model per Section 4.7.3 of the Transition Report.Compliance with NFPA 805 Requirements Page 81 Compliance with NFPA 805 Requirements Page 81 V- RC-11-0149 Section 4.0 4.9 Supplemental Information

-Other VCSNS Specific Issues 4.9.1 Self-Induced Station Blackout (SISBO)The previous Appendix R methodology involves intentionally de-energizing both offsite power and one on-site emergency power sources to prevent spurious operation of equipment.

The SISBO methodology will be eliminated during the NFPA 805 transition process.Background Section III.G of Appendix R to 10 CFR 50 stipulates the requirements to ensure the ability to achieve and maintain safe shutdown conditions.

At least one train of equipment and systems required to achieve and maintain hot shutdown conditions is required to be "free of fire damage" from either the control room or emergency control station. The Appendix R compliance assessment methodology at VCSNS credited operator manual actions to intentionally de-energize power to vital power supplies buses to prevent and/or limit the number of spurious operations that could occur as a result of an Appendix R fire, and local operator manual actions were credited to position or verify position of motor and pneumatic operated valves. These operator actions ideally initiated a SISBO condition.

The SISBO methodology is considered to be a significant contributor to Core Damage Frequency because it involves operator actions to de-energize both offsite power sources and on-site emergency power sources.Resolution Because the premise of the current Appendix R analysis relied heavily on operator manual actions in lieu of identifying safe shutdown cables, performing the detailed circuit analysis and identifying the location of the safe shutdown cables, the SISBO elimination strategy involves determining the extent of fire damage of safe shutdown cables in each fire area. This effort involves performing the detailed cable identification, circuit analysis, and locating cables that are required for the safe shutdown equipment to perform the desired function to bring the VCSNS plant from full power safe and stable conditions (Section 4.2.1.2 of the Transition Report) to cold shutdown conditions.

After identification of nuclear safety systems and equipment, circuit analysis, cable location and instrument tube location, the NSCA fire area by area assessment identified deterministic approaches for Nuclear Safety Performance Criteria compliance.

Where deterministic compliance could not be achieved, station modifications were proposed or a performance-based approach was used, documented in Fire Modeling or Fire Risk Evaluations.

The Fire Risk Evaluation evaluated the risk associated with the variant alternative against the fire risk associated with the deterministically compliant alternative.

The delta risk between these two was compared to the risk acceptance criteria for resolution.

Since NEI 04-02 was written primarily to transition existing fire protection programs, and existing shutdown strategies, SCE&G initiated FAQ 09-0057 to address a direct approach to analysis and compliance directly in alignment with NFPA 805.Compliance with NFPA 805 Requirements Page 82 Compliance with NFPA 805 Requirements Page 82 i-ýýe RC-11-0149 Section 4.0 4.9.2 NFPA 805 Chapter 4 Requirements for Approval The following sections of NFPA 805 Chapter 4 below may not have previous NRC approval of an alternate approach, methods and/or condition which VCSNS considers to be minor variations to the NFPA 805 requirements.

4.2.3.3 (b) -Approval is requested for locations in the plant where twenty feet of separation is required, but intervening combustibles exist. The intervening combustibles are in the form of exposed cable trays.The specific deviation is provided in Attachment X. VCSNS requests NRC approval of this proposed alternative and clarification of the FPP elements.Compliance with NFPA 805 Requirements Page 83 Compliance with NFPA 805 Requirements Page 83

_ ý° RC-11-0149 Section 5.0

5.0 REGULATORY EVALUATION

5.1 Introduction

-10 CFR 50.48 On July 16, 2004 the NRC amended 10 CFR 50.48, Fire Protection, to add a new subsection, 10 CFR 50.48(c), which establishes alternative fire protection requirements.

10 CFR 50.48 endorses, with exceptions, NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants -2001 Edition, as a voluntary alternative for demonstrating compliance with 10 CFR 50.48 Section (b), Appendix R, and Section (f), Decommissioning.

The voluntary adoption of 10 CFR 50.48(c) by VCSNS does not eliminate the need to comply with 10 CFR 50.48(a) and 10 CFR 50, Appendix A, GDC 3, Fire Protection.

The NRC addressed the overall adequacy of the regulations during the promulgation of 10 CFR 50.48(c) (Reference FR Notice 69 FR 33536 dated June 16, 2004, ML041340086)."NFPA 805 does not supersede the requirements of GDC 3, 10 CFR 50.48(a), or 10 CFR 50.48(0. Those regulatory requirements continue to apply to licensees that adopt NFPA 805. However, under NFPA 805, the means by which GDC 3 or 10 CFR 50.48(a) requirements may be met is different than under 10 CFR 50.48(b).

Specifically, whereas GDC 3 refers to SSCs important to safety, NFPA 805 identifies fire protection systems and features required to meet the Chapter I performance criteria through the methodology in Chapter 4 of NFPA 805. Also, under NFPA 805, the 10 CFR 50.48(a)(2)(iii) requirement to limit fire damage to SSCs important to safety so that the capability to safely shut down the plant is ensured is satisfied by meeting the performance criteria in Section 1.5.1 of NFPA 805. The Section 1.5.1 criteria include provisions for ensuring that reactivity control, inventory and pressure control, decay heat removal, vital auxiliaries, and process monitoring are achieved and maintained.

This methodology specifies a process to identify the fire protection systems and features required to achieve the nuclear safety performance criteria in Section 1.5 of NFPA 805. Once a determination has been made that a fire protection system or feature is required to achieve the performance criteria of Section 1.5, its design must meet any applicable requirements of NFPA 805, Chapter 3.Having identified the required fire protection systems and features, the licensee selects either a deterministic or performance-based approach to demonstrate that the performance criteria are satisfied.

This process satisfies the GDC 3 requirement to design and locate SSCs important to safety to minimize the probability and effects of fires and explosions." (Reference FR Notice 69 FR 33536 dated June 16, 2004, ML041340086)

The new rule provides actions that may be taken to establish compliance with 10 CFR 50.48(a), which requires each operating nuclear power plant to have a fire protection program plan that satisfies GDC 3, as well as specific requirements in that section. The transition process described in 10 CFR 50.48(c)(3)(ii) provides, in pertinent parts, that a licensee intending to adopt the new rule must, among other things, "modify the fire protection plan required by paragraph (a) of that section to reflect the licensee's decision to comply with NFPA 805." Therefore, to the extent that the Regulatory Evaluation Page 84 RC-11-0149 Section 5.0 contents of the existing fire protection program plan required by 10 CFR 50.48(a) are inconsistent with NFPA 805, the fire protection program plan must be modified to achieve compliance with the requirements in NFPA 805. All other requirements of 10 CFR 50.48 (a) and GDC 3 have corresponding requirements in NFPA 805.A comparison of the current requirements in Appendix R with the comparable requirements in Section 3 of NFPA 805 shows that the two sets of requirements are consistent in many respects.

This was further clarified in FAQ 07-0032, 10 CFR 50.48(a) and GDC 3 clarification (ML081400292).

The following tables provide a cross reference of fire protection regulations associated with the post-transition VCSNS fire protection program and applicable industry and VCSNS documents that address the topic.10 CFR 50.48(a)Table 5-1 10 CFR 50.48(a) -ApplicabilitylCompliance Reference 10 CFR 50.48(a) Section(s)

ApplicabilitylCompliance Reference (1) Each holder of an operating license issued under this See below part or a combined license issued under part 52 of this chapter must have a fire protection plan that satisfies Criterion 3 of appendix A to this part. This fire protection plan must: (i) Describe the overall fire protection program for the NFPA 805 Section 3.2 facility; NEI 04-02 Table B-1 (ii) Identify the various positions within the licensee's NFPA 805 Section 3.2.2 organization that are responsible for the program; NEI 04-02 Table B-1 (iii) State the authorities that are delegated to each of NFPA 805 Section 3.2.2 these positions to implement those responsibilities; and NEI 04-02 Table B-1 (iv) Outline the plans for fire protection, fire detection NFPA 805 Section 2.7 and Chapters 3 and 4 and suppression capability, and limitation of fire NEI 04-02 B-1 and B-3 Tables damage.(2) The plan must also describe specific features See below necessary to implement the program described in paragraph (a)(1) of this section such as: (i) Administrative controls and personnel requirements NFPA 805 Sections 3.3.1 and 3.4 for fire prevention and manual fire suppression NEI 04-02 Table B-1 activities;(ii) Automatic and manually operated fire detection and NFPA 805 Sections 3.5 through 3.10 and suppression systems; and Chapter 4 NEI 04-02 B-1 and B-3 Tables (iii) The means to limit fire damage to structures, NFPA 805 Section 3.3 and Chapter 4 systems, or components important to safety so that the NEI 04-02 B-3 Table capability to shut down the plant safely is ensured.(3) The licensee shall retain the fire protection plan and NFPA 805 Section 2.7.1.1 requires that each change to the plan as a record until the documentation (Analyses, as defined by NFPA 805 Commission terminates the reactor license. The 2.4, performed to demonstrate compliance with this licensee shall retain each superseded revision of the standard) be maintained for the life of the plant.procedures for 3 years from the date it was See Attachment A.superseded.

Regulatory Evaluation Page 85

-ýo RC-11-0149 Section 5.0 Table 5-1 10 CFR 50.48(a) -Applicability/Compliance Reference 10 CFR 50.48(a) Section(s)

Applicability/Compliance Reference (4) Each applicant for a design approval, design Not applicable VCSNS is licensed under certification, or manufacturing license under part 52 of 10 CFR 50.this chapter must have a description and analysis of the fire protection design features for the standard plant necessary to demonstrate compliance with Criterion 3 of appendix A to this part.General Design Criterion 3 Table 5-2 GDC 3 -Applicability/Compliance Reference GDC 3, Fire Protection, Statement ApplicabilitylCompliance Reference Structures, systems, and components important to NFPA 805 Chapters 3 and 4 safety shall be designed and located to minimize, NEI 04-02 B-1 and B-3 Tables consistent with other safety requirements, the probability and effect of fires and explosions.

Noncombustible and heat resistant materials shall be NFPA 805 Sections 3.3.2, 3.3.3, 3.3.4, 3.11.4 used wherever practical throughout the unit, NEI 04-02 B-1 Table particularly in locations such as the containment and control room.Fire detection and fighting systems of appropriate NFPA 805 Chapters 3 and 4 capacity and capability shall be provided and designed NEI 04-02 B-1 and B-3 Tables to minimize the adverse effects of fires on structures, systems, and components important to safety.Firefighting systems shall be designed to assure that NFPA 805 Sections 3.4 through 3.10 and 4.2.1 their rupture or inadvertent operation does not NEI 04-02 Table B-3 significantly impair the safety capability of these structures, systems, and components Regulatory Evaluation Page 86 Regulatory Evaluation Page 86

'WCEAMMY-RC-11-0149 Section 5.0 10 CFR 50.48(c)Table 5-3 10 CFR 50.48(c) -ApplicabilitylCompliance Reference 10 CFR 50.48(c) Section(s)

ApplicabilitylCompliance Reference (1) Approval of incorporation by reference.

National Fire Protection Association General Information.

NFPA (NFPA) Standard 805, "Performance-Based Standard for Fire Protection for 805 (2001 edition) is the Light Water Reactor Electric Generating Plants, 2001 Edition" (NFPA 805), edition used.which is referenced in this section, was approved for incorporation by reference by the Director of the Federal Register pursuant to 5 U.S.C. 552(a)and 1 CFR part 51.(2) Exceptions, modifications, and supplementation of NFPA 805. As used in General Information.

NFPA this section, references to NFPA 805 are to the 2001 Edition, with the 805 (2001 edition) is the following exceptions, modifications, and supplementation:

edition used.(i) Life Safety Goal, Objectives, and Criteria.

The Life Safety Goal, The Life Safety Goal, Objectives, and Criteria of Chapter 1 are not endorsed.

Objectives, and Criteria of Chapter 1 of NFPA 805 are not part of the LAR.(ii) Plant Damage/Business Interruption Goal, Objectives, and Criteria.

The The Plant Damage/Business Plant Damage/Business Interruption Goal, Objectives, and Criteria of Interruption Goal, Objectives, Chapter 1 are not endorsed.

and Criteria of Chapter 1 of NFPA 805 are not part of the LAR.(iii) Use of feed-and-bleed.

In demonstrating compliance with the Feed and bleed is not utilized performance criteria of Sections 1.5.1(b) and (c), a high-pressure as the sole fire-protected safe charging/injection pump coupled with the pressurizer power-operated relief shutdown methodology.

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 pressurized-water reactors (PWRs) is not permitted.(iv) Uncertainty analysis.

An uncertainty analysis performed in accordance Uncertainty analysis was not with Section 2.7.3.5 is not required to support deterministic approach performed for deterministic calculations.

methodology.(v) Existing cables. In lieu of installing cables meeting flame propagation Electrical cable construction tests as required by Section 3.3.5.3, a flame-retardant coating may be complies with a flame applied to the electric cables, or an automatic fixed fire suppression system propagation test that was may be installed to provide an equivalent level of protection.

In addition, the found acceptable to the NRC italicized exception to Section 3.3.5.3 is not endorsed.

as documented in NEI 04-02 Table B-1.(vi) Water supply and distribution.

The italicized exception to Section 3.6.4 is See Section 4.1.2.2.not endorsed.

Licensees who wish to use the exception to Section 3.6.4 must submit a request for a license amendment in accordance with paragraph (c)(2)(vii) of this section.Regulatory Evaluation Page 87 Regulatory Evaluation Page 87 A-8 RC-11-0149 Section 5.0 Table 5-3 10 CFR 50.48(c) -Applicability/Compliance Reference 10 CFR 50.48(c) Section(s)

Applicability/Compliance Reference (vii) Performance-based methods. Notwithstanding the prohibition in Section The use of performance-

3.1 against

the use of performance-based methods, the fire protection based methods for NFPA 805 program elements and minimum design requirements of Chapter 3 may be Chapter 3 is requested.

See subject to the performance-based methods permitted elsewhere in the Attachment L.standard.

Licensees who wish to use performance-based methods for these fire protection program elements and minimum design requirements shall submit a request in the form of an application for license amendment under §50.90. The Director of the Office of Nuclear Reactor Regulation, or a designee of the Director, may approve the application if the Director or designee determines that the performance-based approach;(A) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release;(B) Maintains safety margins; and (C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).

(3) Compliance with NFPA 805. See below (i) A licensee may maintain a fire protection program that complies with The LAR was submitted in NFPA 805 as an alternative to complying with paragraph (b) of this section accordance with for plants licensed to operate before January 1, 1979, or the fire protection 10 CFR 50.90. The LAR license conditions for plants licensed to operate after January 1, 1979. The included applicable license licensee shall submit a request to comply with NFPA 805 in the form of an conditions, orders, technical application for license amendment under § 50.90. The application must specifications/bases that identify any orders and license conditions that must be revised or needed to be revised and/or superseded, and contain any necessary revisions to the plant's technical superseded.

specifications and the bases thereof. The Director of the Office of Nuclear Reactor Regulation, or a designee of the Director, may approve the application if the Director or designee determines that the licensee has identified orders, license conditions, and the technical specifications that must be revised or superseded, and that any necessary revisions are adequate.

Any approval by the Director or the designee must be in the form of a license amendment approving the use of NFPA 805 together with any necessary revisions to the technical specifications.(ii) The licensee shall complete its implementation of the methodology in The LAR and transition report Chapter 2 of NFPA 805 (including all required evaluations and analyses) summarize the evaluations and, upon completion, modify the fire protection plan required by paragraph and analyses performed in (a) of this section to reflect the licensee's decision to comply with NFPA 805, accordance with Chapter 2 of before changing its fire protection program or nuclear power plant as NFPA 805.permitted by NFPA 805.(4) Risk-informed or performance-based alternatives to compliance with NFPA No risk-informed or 805. A licensee may submit a request to use risk-informed or performance-performance-based based alternatives to compliance with NFPA 805. The request must be in alternatives to compliance the form of an application for license amendment under § 50.90 of this with NFPA 805 (per chapter. The Director of the Office of Nuclear Reactor Regulation, or 10 CFR 50.48(c)(4))

were designee of the Director, may approve the application if the Director or utilized.

See Attachment P.designee determines that the proposed alternatives: (i) Satisfy the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release;(ii) Maintain safety margins; and (iii) Maintain fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).

Regulatory Evaluation Page 88 q -C"* RC-11-0149 Section 5.0 5.2 Regulatory Topics 5.2.1 License Condition Changes The current VCSNS fire protection license condition 2.c (18) is being replaced with the standard license condition based upon Regulatory Position 3.1 of RG 1.205, as shown in Attachment M.5.2.2 Technical Specifications VCSNS conducted a review of the Technical Specifications to determine which Technical Specifications are required to be revised, deleted, or superseded.

VCSNS determined that the changes to the Technical Specifications and applicable justification listed in Attachment N are adequate for the VCSNS adoption of the new fire protection licensing basis.5.2.3 Orders and Exemptions A review was conducted of the VCSNS docketed correspondence to determine if there were any orders or exemptions that needed to be superseded or revised. A review was also performed to ensure that compliance with the physical protection requirements, security orders, and adherence to those commitments applicable to the plant are maintained.

A discussion of affected orders and exemptions is included in Attachment

0.5.3 Regulatory

Evaluations 5.3.1 No Significant Hazards Consideration A written evaluation of the significant hazards consideration of a proposed license amendment is required by 10 CFR 50.92. According to 10 CFR 50.92, a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: " Involve a significant increase in the probability or consequences of an accident previously evaluated; or" Create the possibility of a new or different kind of accident from any accident previously evaluated; or" Involve a significant reduction in a margin of safety.This evaluation is contained in Attachment Q.Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. VCSNS has evaluated the proposed amendment and determined that it involves no significant hazards consideration.

5.3.2 Environmental

Consideration Pursuant to 10 CFR 51.22(b), an evaluation of the LAR has been performed to determine whether it meets the criteria for categorical exclusion set forth in 10 CFR Regulatory Evaluation Page 89

--a RC-11-0149 Section 5.0 51.22(c).

That evaluation is discussed in Attachment R. The evaluation confirms that this LAR meets the criteria set forth in 10 CFR 51.22(c)(9) for categorical exclusion from the need for an environmental impact assessment or statement.

5.4 Transition

Implementation Schedule The following schedule for transitioning VCSNS to the new fire protection licensing basis requires NRC approval of the LAR in accordance with the following schedule: " Implementation of new NFPA 805 fire protection program to include procedure changes, process updates, and training to affected plant personnel.

This will occur one hundred eighty (180) days after NRC approval." Modifications scope and implementation schedule are provided in Attachment S. Appropriate compensatory measures will be maintained until modifications are complete.Regulatory Evaluation Page 90 Regulatory Evaluation Page 90 RC-11-0149 Section 6.0

6.0 REFERENCES

The following references were used in the development of the TR. Additional references are in the NEI 04-02 Tables in the various Attachments.

1. ANSI/ANS-58.23-2007, "American National Standard -Fire PRA Methodology," November 20, 2007.2. ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," February 2, 2009, 3. Federal Register Notice 69 FR 33536, dated June 16, 2004 (ML041340086).
4. Fire Protection Program-Post-Fire Operator Manual Actions, Federal Register, Vol. 71, No. 43, March 6, 2006, pp. 11169-11172.
5. Jones, Walter W., Richard D. Peacock, Glenn P. Forney, Paul A. Reneke, CFAST -Consolidated Model of Fire Growth and Smoke Transport (Version 6), Technical Reference Guide, Special Publication 1026, National Institute of Standards and Technology, Gaithersburg, MD, April 2009.6. Letter, Annette L. Vietti-Cook, Secretary to R. W. Borchardt, Executive Director for Operations, "Staff Requirements

-SECY-1 1-0033 -Proposed NRC Staff Approach to Address Resource Challenges Associated with Review of a Large Number of NFPA 805 License Amendment Requests," April 20, 2011 (ML1 11101452).

7. Letter, Annette L. Vietti-Cook, Secretary to R. W. Borchardt, Executive Director for Operations, "Staff Requirements

-SECY-1 1-0061 -A Request to Revise the Interim Enforcement Policy for Fire Protection Issues on 10 CFR 50.48(C) to Allow Licensees to Submit License Amendment Requests in a Staggered Approach (RIN 3150-AG48)," June 10, 2011 (ML1 116106160).

8. Letter, NRC to NEI, "Process for Frequently Asked Questions For Title 10 of The Code Of Federal Regulations, Part 50.48(c) Transitions," July 12, 2006 (ML061660105).
9. Letter, NRC to SCE&G, "Deviation from 10 CFR Part 50, Appendix R, Section III.G. Fire Protection of Safe Shutdown Capability for Virgil C. Summer Nuclear Station," October 17, 1997 (TAC No. M97337).10. Letter, NRC to SCE&G, "NRC Response To Progress Energy's Letter Of Intent To Adopt 10 CFR 50.48(c) (NFPA 805 Rule)," January 19, 2007 (ML063520409).
11. Letter, NRC to SCE&G, "Evaluation of the Request for an Extension of Enforcement Discretion in Accordance with the Interim Enforcement Policy for Fire Protection Issues During Transition to National Fire Protection Standard NFPA 805," October 19, 2009 (ML092920297).
12. Letter, SCE&G to NRC, "Letter Of Intent to Adopt NFPA 805, Performance-Based Standard For Fire Protection For Light Water Reactor Electric Generating Plants, 2001 Edition," October 19, 2006 (ML062990543).
13. Letter, SCE&G to NRC, "Request for Extension of Enforcement Discretion and Revised Submittal Schedule for 10 CFR 50.48(c) License Amendment Request (LAR 08-03929)," July 16, 2009.References Page 91 RC-11-0149 Section 6.0 14. NEI 00-01, "Guidance for Post-Fire Safe Shutdown Circuit Analysis," Revision 1, January 2005.15. NEI 00-01, "Guidance for Post-Fire Safe Shutdown Circuit Analysis," Revision 2, May 2009.16. NEI 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program under 10 CFR 50.48(c)," Revision 2, April 2008.17. NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition.18. NRC Enforcement Policy, Policy Statement:

Revision, Federal Register, Vol. 69, No. 115, June 16, 2004, pp. 33684-33685.

19. NRC Enforcement Policy: Extension of Discretion Period of Interim Enforcement Policy, Federal Register, Vol. 71, No. 74, April 18, 2006, pp. 19905-19907.
20. NRC Enforcement Policy: Extension of Enforcement Discretion of Interim Policy, Policy Statement:

Revision, Federal Register, Vol. 70, No. 10, January 14, 2005, pp. 2662-2664.

21. NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown Management." 22. NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Revision 3, July 2000.23. NUREG-1805, Fire Dynamics Tools (FDTs). U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, DC: 2004.24. NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities," April 2005.25. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 1, November 2002.26. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, March 2009.27. Regulatory Guide 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1, December 2009.28. Regulatory Information Summary 2006-10, "Regulatory Expectations with Appendix R Paragraph III.G.2 Operator Manual Actions," June 30, 2006.29. Regulatory Information Summary 2007-19, "Process For Communicating Clarifications Of Staff Positions Provided In Regulatory Guide 1.205 Concerning Issues Identified During The Pilot Application of NFPA 805," August 20, 2007.30. Safety Evaluation Report related to the operation of Virgil C. Summer Nuclear Station, Unit. No. 1, Docket No. 50-395, February, 1981.31. Safety Evaluation Report related to the operation of Virgil C. Summer Nuclear Station, Unit. No. 1, Docket No. 50-395, Supplement No. 2, May, 1981.32. Safety Evaluation Report related to the operation of Virgil C. Summer Nuclear Station, Unit. No. 1, Docket No. 50-395, Supplement No. 3, January, 1982.References Page 92 References Page 92 RC-11-0149 Section 6.0 33. Safety Evaluation Report related to the operation of Virgil C. Summer Nuclear Station, Unit. No. 1, Docket No. 50-395, Supplement No. 4, August, 1982.34. SECY-03-0100, "Rulemaking Plan on Post-Fire Operator Manual Actions," June 17, 2003.35. SECY-06-0010, "Withdraw Proposed Rulemaking

-Fire Protection Program Post-Fire Operator Manual Actions," January 12, 2006.36. SECY-1 1-0033, "Proposed NRC Staff Approach to Address Resource Challenges Associated with Review of a Large Number of NFPA 805 License Amendment Requests," March 4, 2011.37. SECY-1 1-0061, "A Request to Revise the Interim Enforcement Policy for Fire Protection Issues on 10 CFR 50.48(C) to Allow Licensees to Submit License Amendment Requests in a Staggered Approach (RIN 3150-AG48)," April 29, 2011.38. Voluntary Fire Protection Requirement for Light-Water Reactors; Adoption of NFPA 805 as a Risk-Informed, Performance-Based Alternative, Final Rule, Federal Register, Vol. 69, No. 115, June 16, 2004, pp. 33536-33551.

39. Voluntary Fire Protection Requirements for Light-Water Reactors; Adoption of NFPA 805 as a Risk-Informed, Performance-Based Alternative, Proposed Rule, Federal Register, Vol. 67, No. 212, November 1,2002, pp. 66578-66588.

References Page 93 References Page 93 ANRC-11-0149 ATTACHMENTS Attachments Page 94 Attachments Page 94 4rw-4wWg:p-dffrGO RC-11-0149 Attachment A A. NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements 45 Pages Attached NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-1 q-°. , RC-11-0149 Attachment A Transition of Fundamental Fire Protection Program and Design Elements Each section and subsection of NFPA 805 Chapter 3 was reviewed against the current station fire protection program. Upon completion of the activities associated with the review, one or more of the following compliance statement(s) were used: " Complies (C) -The existing FPP elements are determined to meet the requirements of NFPA 805 Chapter 3 element.Acknowledgement and/or restatement of the requirement are not required.

An open item in this category means there are action items to be completed during implementation prior to transition.

Complies directly with the requirements of NFPA 805 Chapter 3." Complies by Alternative (CA) -The existing FPP elements meet the requirements of NFPA 805 by using clarification and/or equivalent alternative(s).

VCSNS requests NRC review/approval of those CA items listed in Section 4.1.2.3 (Table 4-1) of the Transition Report and included in Attachment L. Complies with clarification with the requirements of NFPA 805 Chapter 3." Complies with Fire Protection Engineering Equivalency Evaluations (CE) -The existing FPP elements have been determined to be adequate for the hazard by a FPE and to meet the NFPA 805 Chapter 3 requirements.

Complies through the use of Fire Protection Engineering Equivalency Evaluations (FPEEE) which are valid and of appropriate quality. VCSNS requests NRC review/approval of those Engineering Evaluations listed in Section 4.1.2.3 (Table 4-1) of the Transition Report and included in Attachment L." Complies by Previous NRC Approval (CNRC) -The existing FPP elements specified in NFPA 805 Chapter 3 requirements are not in strict compliance, however, previous NRC approval of the configuration exists. An NRC approved alternative or deviation to NFPA 805 Chapter 3, would supplant the specific requirement of NFPA 805 Chapter 3. Where credited, these prior approvals have been incorporated into an FPEEE, and included in Attachment K and Attachment L." No Review Required (NRR) -The existing Chapter 3 elements are not based on the requirements and/or are not applicable to elements of the VCSNS Fire Protection Program.NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-2 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary This chapter contains the fundamental elements of the fire protection program and specifies the minimum design requirements for the fire protection systems and features.These fire protection program elements and minimum design requirements shall not be subject to the performance-based methods permitted elsewhere in this standard.Previously approved alternatives from the fundamental protection program attributes of this chapter by the AHJ take precedence over the reauirements contained herein.NRR Section heading with no requirements requiring evaluation.

See subsections for requirements.

3.2.1 Intent. A site-wide fire protection plan shall be established.

This plan shall document management policy and program direction and shall define the responsibilities of those individuals responsible for the plan's implementation.

This section establishes the criteria for an integrated combination of components, procedures, personnel to implement all fire protection activities.

C SAP-0131, "Fire Protection Program", Rev 6D The Station Administrative Procedure, SAP-131 defines and describes the Fire Protection Program (FPP) including responsibilities, program elements, and procedures to ensure effective implementation.

The regulatory basis for this FPP is 10 CFR 50.48 Criterion 3 of Appendix A to this part, including Appendix R and NFPA 805-2001. (Table S-2, Item 1)3.2.2 Management Policy Direction and Responsibility A policy document shall be prepared that C SAP-0131, "Fire Protection The Station Administrative Procedure (SAP)defines management authority and Program", Rev 6D, is consistent with other upper tier policy responsibilities and establishes the general positions/

program documents at VCSNS. It policy for the site fire protection program. provides a level of authority and responsibility for all groups and organizations for interfaces with the Fire Protection Program (FPP).NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-3 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary 3.2.2.1 The policy document shall designate the C SAP-0131, "Fire Protection The Station Administrative Procedure (SAP)senior management position with immediate Program", Rev 6D, Section 5.1 defines the General Manager, Nuclear Plant authority and responsibility for the fire Operation as the management position with protection program. ultimate responsibility for the FP Program.3.2.2.2 The policy document shall designate a C SAP-0131, "Fire Protection The Station Administrative Procedure (SAP)position responsible for the daily administration Program", Rev 6D, Section 5.2 defines the Fire Protection Coordinator as the and coordination of the fire protection program position responsible for the daily administration and its implementation.

of the Fire Protection Program. (Table S-2, Item 1)3.2.2.3 The policy document shall define the fire C SAP-0131, "Fire Protection The Station Administrative Procedure (SAP)protection interfaces with other organizations Program", Rev 60, Section 6.2 defines the interfaces, responsibilities and and assign responsibilities for the coordination authorities for the various elements of the FP of activities.

In addition, this policy document program. (Table S-2, Item 1)shall identify the various plant positions having the authority for implementing the various areas of the fire protection program.3.2.2.4 The policy document shall identify the C SAP-0131, "Fire Protection The Station Administrative Procedure (SAP)appropriate AHJ for the various areas of the Program", Rev 6D defines the AHJ for the various areas of the FP fire protection program. program. (Table S-2, Item 1)3.2.3 Procedures Procedures shall be established for C SAP-0131, "Fire Protection SAP-1 31 is the primary document that implementation of the fire protection program. Program", Rev 6D, Section 6.2 establishes the elements for implementing the In addition to procedures that could be fire protection program, including references to required by other sections of the standard, the the applicable FPP implementing procedures.

procedures to accomplish the following shall be established:

NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-4 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-4 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary (1) Inspection, testing, and maintenance CA SAP-0131, "Fire Protection Inspection, Testing and Maintenance for the for the protection systems and features Program", Rev 6D, Section Fire Protection Program are established in credited by the fire protection program 6.2.4, 6.2.5, 6.2.6 accordance with controlled procedures.

As FPP-023, "Fire Detection", Rev practical, performance-based surveillance 3A Enclosure

6.1 frequencies

may be established as referenced FPP-024, "Fire Suppression", herein and described in Electric Power Rev 3C Enclosure

6.1 Research

Institute (EPRI) technical report and FPP-025, "Fire Containment", NEIL Appendix 4.2.8. (Table S-2, Item 2)Rev 4F Enclosure 6.5, 6.6 EPRI Technical Report (TR)1006756 Fire Protection Surveillance Optimization and Maintenance Guide NEIL Appendix 4.2.8 (2) Compensatory actions implemented C FPP-023, "Fire Detection", Existing procedures address current when fire protection systems and other Rev 3A Enclosure

6.1 compensatory

measures for the program, systems credited by the fire protection FPP-024, "Fire Suppression", however a new procedure establishing revised program and this standard cannot perform Rev 3C Enclosure 6.1 and updated compensatory measures will be their intended function and limits on FPP-025, "Fire Containment", developed during the implementation period to impairment duration Rev 4F Enclosure 6.5, 6.6 incorporate NFPA 805 insights. (Table S-2, Item 1)(3) Reviews of fire protection program- C SAP-0131, "Fire Protection Station Administrative Procedure (SAP)related performance and trends Program", Rev 6D establishes responsibilities for review of ES091 1, "Fire Protection program related performance and trends.Monitoring Program" Rev 0 System Engineering procedures establishes FP feature trends and monitoring. (Table S-2, Item 1) (Table S-2, Item 4)(4) Reviews of physical plant modifications C SAP-0133, "Design Control Engineering Services and Document Control and procedure changes for impact on the fire Program", Rev 14B Procedures manage interfaces and direct protection program SAP-0139, "Document Review appropriate documents to appropriate and Approval Process", Rev 32 personnel for FP program impacts. (Table S-2, ES-427 "Program / Issues Item 3)Screening" Rev 2D NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-5 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-5

-ý RC-11-0149 Attachment A RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary (5) Long-term maintenance and C MD-21 "Configuration Long term maintenance and configuration of configuration of the fire protection program Management", Rev 7 the fire protection program is implemented SAP-1 31 "Fire Protection through a Management Directive (MD) and Program", Rev 6D Station Administrative Procedures (SAP).SAP-0139 "Document Review and Approval Process", Rev 32 SAP-0133, "Design Control Program", Rev 14B (6) Emergency response procedures for C EPP-013 "Fire Emergency", Station Emergency Response Procedures the plant industrial fire brigade Rev 8 direct plant actions for plant fire brigade response due to fires at the station.A fire prevention program with the goal of NRR FPP022 "Fire Prevention", Elements of the fire prevention program are preventing a fire from starting shall be Rev 3 described in the following subsections.

established, documented, and implemented as part of the fire protection program. The two basic components of the fire prevention program shall consist of both of the following:

(1) Prevention of fires and fire spread by C SAP0131 "Fire Protection FP Program impacts due to expected controls on operational activities Program", Rev 14B operational activities are described in the Fire FPP022 "Fire Prevention", Rev 3 Protection Program, and implemented in FPP020 "Program described station programs. (Table S-2, Item 1)Administration", Rev 5E (2) Design controls that restrict the use of C SAP-0131, "Fire Protection The control of fixed and transient materials combustible materials Program", Rev 6D at the station are incorporated into a variety of SAP-142 "Station Housekeeping controlled procedures, to manage the Program" Rev 15B introduction and use of combustible materials SAP133 "Design Control", Rev at the station. (Table S-2, Item 1)14B FPP022 "Fire Prevention", Rev 3 Enclosure 6.2 NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-6 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-6

-ýý RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary The design control requirements listed in the NRR These control elements are described in remainder of this section shall be provided as summary in the following elements of the code described, and are documented, and implemented as part of the fire protection program.3.3.1 Fire Prevention for Operational Activities The fire prevention program activities shall C FPP022 "Fire Prevention", A variety of Fire Protection and Training consist of the necessary elements to address Rev 3 Program procedures manage ignition sources, the control of ignition sources and the use of FPP020 "Program combustible material and personnel response transient combustible materials during all Administration", Rev 5E should a fire occur. (Table S-2, Item 1)aspects of plant operations.

The fire TQP-606 "General Employee prevention program shall focus on the human Training Fire Protection Training" and programmatic elements necessary to Rev 1A prevent fires from starting or, should a fire start, to keep the fire as small as possible.3.3.1.1 General Fire Prevention Activities The fire prevention activities shall include NRR Individual elements are addressed below, but not be limited to the following program but not limited to, the identified elements.elements: (1) Training on fire safety information for all C TQP-606 "General Employee The Fire Protection Training Program for all employees and contractors including, as a Training Fire Protection Training" employees and contractors is designed to minimum, familiarization with plant fire Rev 1A familiarize personnel with their responsibilities prevention procedures, fire reporting, and plant associated with fire events at the station.emergency alarms. (Table S-2, Item 4) (Table S-2, Item 17)(2)

  • Documented plant inspections including C QSP-208 "Inspection of Plant inspections are conducted in provisions for corrective actions for conditions Housekeeping and Items in accordance with a variety of station where unanalyzed fire hazards are identified.

Storage", Rev 14 procedures.

Corrective Actions are specified SAP142 "Station Housekeeping within the scope of the procedure.

Program", Rev 15B QSP-1 06 "Conduct of Quality Assurance Activities", Rev 17B VCSNS Technical Specifications-Audits Sections 6.5.2.8 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-7 RC-11-0149 Attachment A NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary (3)

  • Administrative controls addressing the C SAP-0131, "Fire Protection Oversight of potential impacts on the Fire review of plant modifications and maintenance Program", Rev 6D Protection Program, which includes the to ensure that both fire hazards and the impact SAP133 "Design Control", Rev identification of fire hazards and potential on plant fire protection systems and features 14B impacts on systems/ features are controlled.

are minimized.

SSP-001 "Planning and (Table S-2, Item 15) (Table S-2, Item 16)Scheduling Maintenance Activities", Rev 22 SSP-002 "Planning and Scheduling of Outage Maintenance Activities" (PSE), Rev 6 ES-427 "Program Issue Screening" Rev 2D 3.3.1.2 Control of Combustible Materials 3.3.1.2 Procedures for the control of general C QSP-208 "Inspection of Individual elements are addressed below, housekeeping practices and the control of Housekeeping and items in but not limited to these elements.transient combustibles shall be developed and Storage", Rev 14 implemented.

These procedures shall include SAP1286 "Procurement of but not be limited to the following program Materials" Rev 7B elements:

SAP-142 "Station Housekeeping Program" Rev 15B (1)

  • Wood used within the power block shall CA FPP022 "Fire Prevention", Station procedures limit the type and use of be listed pressure-impregnated or coated with Rev 3 wood at the station to that described.

The a listed fire-retardant application.

Exception:

TRP-02, "Fire Protection", Rev procurement of wood is incorporated as Fire Cribbing timbers 6 in. by 6 in. (15.2 cm by 15.2 8D Protection Program input into the procurement cm) or larger shall not be required to be fire- process.retardant treated.Untreated wood or lumber will be addressed for controls of limited duration, when required, as part of the fire protection program (e.g.outages, compensatory measures)(Table S-2, Item 1)NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-8 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-8 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary (2) Plastic sheeting materials used in the C FPP022 "Fire Prevention", Station procedures limit the type and use of power block shall be fire-retardant types that Rev 3 Enclosure

6.2 plastic

sheeting materials at the station, to have passed NFPA 701, Standard Methods of TRP-02, "Fire Protection", Rev those with qualified test methods as described.

Fire Tests for Flame Propagation of Textiles 8D Information concerning restrictions concerning and Films, large-scale tests, or equivalent, the procurement plastic sheeting has been addressed as Fire Protection inputs to the procurement process. (Table S-2, Item 1)(3) Waste, debris, scrap, packing materials, C SAP-0300 "Conduct of Station Maintenance procedures require or other combustibles shall be removed from Maintenance", Rev 12C, Section area cleanup following completion of an area immediately following the completion 6.1.17 maintenance activities. (Table S-2, Item 1)of work or at the end of the shift, whichever SAP-0142 "Station comes first. Housekeeping Program", Rev 15B (4)

  • Combustible storage or staging areas C SAP-0142 "Station Station administrative procedures control the shall be designated, and limits shall be Housekeeping Program", Rev designation and management of Combustible established on the types and quantities of 15B Material Storage and Staging Areas (Table S-stored materials.

FPP022 "Fire Prevention", Rev 3 2, Item 1)Enclosure 6.2 (5)

  • Controls on use and storage of C SAP-0131, "Fire Protection Station Administrative and Fire Protection flammable and combustible liquids shall be in CE Program", Rev 6D procedures control the use and storage of accordance with NFPA 30, Flammable and SAP-0403 "Chemical Control Flammable/

Combustible Liquids.Combustible Liquids Code, or other applicable Program", Rev 8 Programmatic controls have been evaluated NFPA standards.

FPP022 "Fire Prevention", Rev 3 against the requirements of NFPA30.Enclosure 6.2 DC0780D-006 "Administrative No other standards, other than NFPA30, are and Program Controls" applicable to the use and storage of Flammable and Combustible Liquids at the station. (Table S-2, Item 1)NEI 04-02 Table 8-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-9 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-9 4r0-4=T4:ffWffEG6 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary (6)

  • Controls on use and storage of C ISP-001 "Administration of Per FAQ 06-0020, the station maintains flammable gases shall be in accordance with CE Safety Program", Rev 8B administrative controls for compressed gases applicable NFPA standards.

FPP022 "Fire Prevention", Rev 3 in accordance with the original NRC guidance DC0780D-006 "Administrative provided by the 'Nuclear Plant Fire Protection and Program Controls" Functional Responsibilities, Administrative Controls, and Quality Assurance' dated June 14, 1977 with the NRC's evaluation in NUREG 0717 "Safety Evaluation Report related to the operation of Virgil C Summer Nuclear Station, Unit No. 1", dated February 1981 (page 9-43).No NFPA standards were determined to be applicable at the time.Bulk storage of flammable gases has been evaluated against the requirements of NFPA 50A-73, "Gaseous Hydrogen Systems at Consumer Sites" (Table S-2, Item 1)3.3.1.3 Control of Ignition Sources 3.3.1.3.1" A hot work safety procedure shall be C FPP022 "Fire Prevention", Administrative controls have been developed developed, implemented, and periodically CE Rev 3 and implemented to permit, and manage Hot updated as necessary in accordance with DC0780D-006 "Administrative Work Permits.NFPA 51B, Standard for Fire Prevention and Program Controls" During Welding, Cutting, and Other Hot Work, Programmatic controls have been evaluated and NFPA 241, Standard for Safeguarding against the requirements of NFPA 51 B and Construction, Alteration, and Demolition applicable NFPA 241 criteria. (Table S-2, Item Operations.

1 )3.3.1.3.2 Smoking and other possible sources of C FPP022 "Fire Prevention", Smoking at the station is restricted to ignition shall be restricted to properly Rev 3 approved locations.

Other sources of ignition designated and supervised safe areas of the MD-64 "Smoking Policy- are controlled through Hot Work Permit.plant. Personnel Located Within The (Table S-2, Item 1)Nuclear Strategic Business Unit" Rev 7 NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-b NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-1 0 RC-11-0149 Attachment A NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary 3.3.1.3.3 Open flames or combustion-generated C FPP022 "Fire Prevention", FP administrative procedures prohibit the smoke shall not be permitted for leak or air Rev 3 use of open flame or combustion-generated flow testing. smoke for use in leak and air flow testing (Table S-2, Item 1)3.3.1.3.4*

Plant administrative procedure shall control C FPP022 "Fire Prevention", FP administrative procedures manage the the use of portable electrical heaters in the Rev 3 use, and installation of portable heaters in the plant. Portable fuel-fired heaters shall not be plant consistent with that described in this permitted in plant areas containing equipment section. (Table S-2, Item 1)important to nuclear safety or where there is a potential for radiological releases resulting from a fire.3.3.2 Structural Walls, floors, and components required to C Drawings 400 Series The structural members of buildings are maintain structural integrity shall be of "Concrete" constructed of non-combustible or limited noncombustible construction, as defined in Drawings 500 Series "Structural combustible materials.

Structural and concrete NFPA 220, Standard on Types of Building Steel" station drawings provide applicable Construction.

Drawings 100 Series construction details."Architectural" Drawing DC0780D-007 "General Station Construction Features &Materials" 3.3.3 Interior Finishes Interior wall or ceiling finish classification C SAP-0131, "Fire Protection Station changes and materials are reviewed shall be in accordance with NFPA 101, Life CE Program", Rev 6D for limitations imposed by Interior finishes as Safety Code, requirements for Class A DC0780D-009 "Life Safety" described in NFPA 101.materials.

Interior floor finishes shall be in accordance with NFPA 101 requirements for Programmatic controls associated with interior Class I interior floor finishes.

finishes have been evaluated against the requirements of NFPA 101, Life Safety Code.(Table S-2, Item 3)3.3.4 Insulation Materials NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-1 I

=-_o__ RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary Thermal insulation materials, radiation C SP-138, "Insulation Outside Thermal insulating materials are heat shielding materials, ventilation duct materials, Containment", Rev 3, Section resistant and non-combustible or limited and soundproofing materials shall be 5:03 combustible materials.

There are no noncombustible or limited combustible.

SP-136, "Insulation Inside soundproofing materials.

SP138 references Containment", Rev 3 ASTM C547 & C553 for mineral wool and SP-424, "Insulation Outside SP424 requires fiberglass blankets.

Ventilation Containment" Rev 0 duct materials are mineral fiber (SP424).Radiation shielding materials are non combustible or limited combustible (Table S-2, Item 3)3.3.5 Electrical 3.3.5.1 Wiring above suspended ceiling shall be CE SP-222 "Electrical Station specifications and procedures govern kept to a minimum. Where installed, electrical Installation", Rev 16 the installation wiring above suspended wiring shall be listed for plenum use, routed in TR0780E-004 "Administrative ceilings, which is kept to a minimum. These armored cable, routed in metallic conduit, or Features & Materials:

Electrical specifications and procedures require wire and routed in cable trays with solid metal top and Wiring and Cabling", Rev 0 cable installed in NFPA 805 credited areas to bottom covers. EMP-391.001: "Installation of be qualified to IEEE 383 flame test or be conduit", Rev 7 plenum rated.EMP-300.003: "Installation of flexible conduit" Rev 13B Future wiring in this type space for the Power SP-371 "Communication Cable", Block are listed for plenum use, routed in Rev 0 armored cable, routed in metallic conduit or SP-372 "Lighting Cable", Rev 0 routed in cable trays with solid metal top and bottoms covers. (Table S-2, Item 3)3.3.5.2 Only metal tray and metal conduits shall be C SP-222 "Electrical Raceways including cable trays and conduit used for electrical raceways.

Thin wall metallic Installation", Rev 16 are constructed of metal tray or conduit. Thin tubing shall not be used for power, Drawing 215-001, Rev 25 wall metallic tubing is not used for power, instrumentation, or control cables. Flexible SP- 558 "Cable Tray", instrumentation or control cables. When used, metallic conduits shall only be used in short Addendum A flexible metallic conduits are only used in lengths to connect components.

Drawing 214-001, Rev 15 limited lengths.Drawing 215-002, Sheet 2, Rev 16 3.3.5.3* Electric cable construction shall comply with CE SP-222 "Electrical Electrical cable insulation of all purpose a flame propagation test as acceptable to the Installation", Rev 16 cable is qualified by flame propagation testing NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-12 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary AHJ. SP-371 "Communication Cable", acceptable to the AHJ (FAQ06-022).

Rev 0 Engineering documents control cable SP-372 "Lighting Wire" Rev 0 construction to comply with the subject SP-374 "Paging System" Rev 0 attributes.

Very small amounts of untested SP-1 511 "Procurement and special purpose cable are installed but do not Installation of Rockbestos result in a significant fire risk (Table S-2, Item Firezone R Cable" Rev 0 3)TR0780E-004 "FP Admin Features & Program Controls", Recognition of alternative flame propagation Rev 0 test methods (FAQ06-022) is captured in the evaluation for future station reference/

configuration control.3.3.6 Roofs Metal roof deck construction shall be CA SP-1 12 "Plant Enclosures", Metal deck roof construction is designed and installed so the roofing system Rev 0 Drawing noncombustible and is listed as Class I by the will not sustain a self-propagating fire on the SP-152 "Roof Insulation, Built Factory Mutual System Approval Guide. (FM underside of the deck when the deck is heated Up Roofing and Sheet Metal" Global Property Loss Prevention Data Sheets by a fire inside the building.

Roof coverings Rev 3 1-31, Metal Roof Systems) or Class A (NFPA shall be Class A as determined by tests DC0780D-006 "Administrative 256). (Table S-2, Item 3)described in NFPA 256, Standard Methods of and Program Controls" Fire Tests of Roof Coverings.

3.3.7 Bulk Flammable Gas Storage Bulk compressed or cryogenic flammable C FPP022 "Fire Prevention", Bulk compressed or cryogenic flammable gas storage shall not be permitted inside Rev 3 gases are stored outside station structures structures housing systems, equipment, or SAP-133 "Design Control", Rev (Table S-2, Item 1)components important to nuclear safety. 14B 3.3.7.1 Storage of flammable gas shall be located C FPP022 "Fire Prevention", Flammable gases are stored outside station outdoors, or in separate detached buildings, so CE Rev 3 structures or in separate detached buildings.

that a fire or explosion will not adversely DC0780D-006 "Administrative impact systems, equipment, or components Features & Materials", Rev 0 Bulk storage of flammable gases has been important to nuclear safety. NFPA 50A, evaluated against the requirements of NFPA Standard for Gaseous Hydrogen Systems at 50A-73, "Gaseous Hydrogen Systems at Consumer Sites, shall be followed for Consumer Sites" hydrogen storage. (Table S-2, Item 1)NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-1 3 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary 3.3.7.2 Outdoor high-pressure flammable gas C Drawing 011-001 Bulk high pressure flammable storage storage containers shall be located so that the CE "Transformer Area" Rev 11 containers are normally located such that the long axis is not pointed at buildings.

TR0780E-006 "Fire Protection long axis is parallel to site structures.

Admin and Program Controls" The generator hydrogen storage tank south of the Turbine Building is perpendicular to the Turbine Building, but spatially separated from the building (>200 feet) and discussed in the engineering evaluation.

3.3.7.3 Flammable gas storage cylinders not C FPP022 "Fire Prevention", When not in use portable compressed gas required for normal operation shall be isolated Rev 3 cylinders are isolated (Table S-2, Item 1)from the system. WM-3.0 "Welding Safety" Rev 6B, Section 5.1.10 3.3.8 Bulk Storage of Flammable and C FPP022 "Fire Prevention", Administrative procedures prohibit the bulk Combustible Liquids Bulk storage of CE Rev 3 storage of flammable and combustible liquids flammable and combustible liquids shall not be PTP-1 14.091 "Flammable Liquid inside site structures.

permitted inside structures containing systems, Locker Inspection", Rev 3C equipment, or components important to SAP-142 "Station Housekeeping Bulk storage of flammable and combustible nuclear safety. As a minimum, storage and Program" Rev 15B liquids has been evaluated against the use shall comply with NFPA 30, Flammable DC0780D-006 "Administrative requirements of NFPA 30, "Flammable and and Combustible Liquids Code. and Program Controls" Combustible Liquids Code" (Table S-2, Item 1)3.3.9* Transformers Where provided, transformer C PTP 114.073 "Transformer Visual inspections are performed during oil collection basins and drain paths shall be Deluge Operational Test" Rev 6 periodic transformer water spray testing to periodically inspected to ensure that they are TR07800-013 "Transformer assess collection basis and drain path free of debris and capable of performing their Hazards Analysis SOER1 0-1", performance.

design function.

Rev 0 A Fire Hazard Evaluation of the Transformer area considered drainage alternatives to that cited in this section. (Table S-2, Item 2)NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-14 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-14 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary 3.3.10 Hot Pipes and Surfaces.

Combustible C FPP022 "Fire Prevention", Administrative procedures ensure the prompt liquids, including high flashpoint lubricating Rev 3 identification and correction of any combustible oils, shall be kept from coming in contact with SAP-142 "Station Housekeeping liquid leakage at the station, which would hot pipes and surfaces, including insulated Program" Rev 15B include hot pipes and surfaces.

This would pipes and surfaces.

Administrative controls QSP-208 "Inspection of include housekeeping considerations with the shall require the prompt cleanup of oil on Housekeeping and Items In use of combustible liquids during maintenance insulation.

Storage", Rev 14 periods. (Table S-2, Item 1)SAP1256 "Leak Reduction Program", Rev 1 3.3.11 Electrical Equipment.

Adequate C FPP022 "Fire Prevention", Placement of combustible materials in clearance, free of combustible material, shall Rev 3 proximity to energized electrical equipment is be maintained around energized electrical QSP-208 "Inspection of controlled. (Table S-2, Item 1)equipment.

Housekeeping and Items In Storage", Rev 14, SAP-142 "Station Housekeeping Program" Rev 15B 3.3.12 Reactor Coolant Pumps. For facilities with C Drawing 305-601 Sheet 1, The RCP oil collection system has been non-inerted containments, reactor coolant Reactor Coolant Pump A Oil designed, engineered and installed that failure pumps with an external lubrication system Collection Systems Rev 1 will not lead to fire during normal or design shall be provided with an oil collection system. Drawing 305-601 Sheet 2, basis accident conditions and that there is The oil collection system shall be designed Reactor Coolant Pump B Oil reasonable assurance that the system will and installed such that leakage from the oil Collection Systems Rev 1 withstand the Safe Shutdown Earthquake (10 system is safely contained for off normal Drawing 305-601 Sheet 3, CFR 50, Appendix R,Section III 0). The conditions such as accident conditions or Reactor Coolant Pump C Oil system meets the five criteria presented in this earthquakes.

All of the following shall apply: Collection Systems Rev 1 section. A stress margin exists for the pump 302-606 "RCP Oil Collection enclosures, the drain piping, and the collection System" Rev 1 tank in that the stress ratios during an SSE are FR DBD "10CFR50 Appendix less than 1.00. Components of the oil R", Rev 4F, Section 4.1 collection system could survive earthquakes of greater magnitude than that postulated in the analysis.(a) The oil collection system for each reactor C Drawing 305-601 Sheet 1, The RCP oil collection system is designed to coolant pump shall be capable of collecting Reactor Coolant Pump A Oil collect oil from pressurized and non lubricating oil from all potential pressurized Collection Systems Rev 1 pressurized leakage sites.and non pressurized leakage sites in each Drawing 305-601 Sheet 2, NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-1 5 4lrw4W4:F--&rG*

RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary reactor coolant pump oil system. Reactor Coolant Pump B Oil Collection Systems Rev 1 Drawing 305-601 Sheet 3, Reactor Coolant Pump C Oil Collection Systems Rev 1 302-606 "RCP Oil Collection System" Rev 1 FR DBD "10CFR50 Appendix R", Rev 4F, Section 4.1 (b) Leakage shall be collected and drained C Drawing 305-601 Sheet 1, The RCP oil collection system is designed to to a vented closed container that can hold the Reactor Coolant Pump A Oil collect leakage, and drain to a vented closed inventory of the reactor coolant pump Collection Systems Rev 1 container, sized to hold the contents of the lubricating oil system. Drawing 305-601 Sheet 2, RCP lubricating system. The individual tank Reactor Coolant Pump B Oil capacities of 275 gallons account for any pump Collection Systems Rev 1 overfill or tank condensation which could occur.Drawing 305-601 Sheet 3, Reactor Coolant Pump C Oil Collection Systems Rev 1 302-606 "RCP Oil Collection System" Rev 1 FR DBD "10CFR50 Appendix R", Rev 4F, Section 4.1 (c) A flame arrestor is required in the vent if C Drawing 305-601 Sheet 1, A flame arrestor has been installed on the the flash point characteristics of the oil present Reactor Coolant Pump A Oil vent for the RCP oil drainage tank.the hazard of a fire flashback.

Collection Systems Rev 1 Drawing 305-601 Sheet 2, Reactor Coolant Pump B Oil Collection Systems Rev 1 Drawing 305-601 Sheet 3, Reactor Coolant Pump C Oil Collection Systems Rev 1 302-606 "RCP Oil Collection System" Rev 1 FR DBD "10CFR50 Appendix R", Rev 4F, Section 4.1 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-1 6 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary (d) Leakage points on a reactor coolant C Drawing 305-601 Sheet 1, The RCP oil collection system is designed to pump motor to be protected shall include but Reactor Coolant Pump A Oil encompass the defined potential leakage not be limited to the lift pump and piping, Collection Systems Rev 1 points for the oil lubrication system.overflow lines, oil cooler, oil fill and drain lines Drawing 305-601 Sheet 2, and plugs, flanged connections on oil lines, Reactor Coolant Pump B Oil and the oil reservoirs, where such features Collection Systems Rev 1 exist on the reactor coolant pumps. Drawing 305-601 Sheet 3, Reactor Coolant Pump C Oil Collection Systems Rev 1 (e) The collection basin drain line to the C Drawing 305-601 Sheet 1, The RCP oil collection system basin drain collection tank shall be large enough to Reactor Coolant Pump A Oil line is sized to accommodate the largest accommodate the largest potential oil leak Collection Systems Rev 1 potential oil leak. ECR-50371

& Appendix R such that oil leakage does not overflow the Drawing 305-601 Sheet 2, DBD evaluates the potential spill into the basin.basin. Reactor Coolant Pump B Oil Collection Systems Rev 1 Drawing 305-601 Sheet 3, Reactor Coolant Pump C Oil Collection Systems Rev 1 302-606 "RCP Oil Collection System" Rev 1 FR DBD "1 OCFR50 Appendix R", Rev 4F, Section 4.1 3.4.1 On-Site Fire-Fighting Capability (a) A fully staffed, trained, and equipped fire- C SAP-131 "Fire Protection A fully staffed, trained and equipped five (5)fighting force shall be available at all times to Program", Rev 6D man fire brigade is available at all times to control and extinguish all fires on site. This TQP-606 "General Employee respond to, control and extinguish fires on site.force shall have a minimum complement of Training Fire Protection five persons on duty and shall conform with Training", Rev 1A the following NFPA standards as applicable:

Drawing DC0780D-009"Industrial Fire Brigade" OAP1 00.6" Control Room Conduct and Control of Shift Activities" Attachment VIIA Rev NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-17 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-1 7 wn--ý41r4%1:K"6 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary 2E (1) NFPA 600, Standard on Industrial Fire CE EPP-107 "Fire Brigade", Rev The station fire brigade has been evaluated Brigades (Interior Structural Fire Fighting)

OA against the requirements of NFPA 600, TQP-606 "General Employee "Standard on Industrial Fire Brigades" Training Fire Protection Training", Rev 1A DC0780D-008 Industrial Fire Brigade, Rev 0 (2) NFPA 1500, Standard on Fire NRR NFPA600 is used for the Industrial Fire Department Occupational Safety and Health Brigade.Program (3) NFPA 1582, Standard on Medical NRR FAQ 06-007 "Clarification on This standard applies to fire department Requirements for Fire Fighters and Information Plant Fire Brigades" organizations only. VCSNS is organized as a for Fire Department Physicians DC0780D-008 Industrial Fire fire brigade.Brigade, Rev 0 (b)

  • Industrial fire brigade members shall C EPP-107 "Fire Brigade", Rev During an event requiring fire brigade have no other assigned normal plant duties OA response (e.g. station fire), the assigned fire that would prevent immediate response to a DC0780D-008 Industrial Fire brigade members primary responsibility is to fire or other emergency as required.

Brigade, Rev 0 support timely response/

resolution to the event.(c) During every shift, the brigade leader and C SAP-200 "Conduct of At the start of every shift, qualifications of the at least two brigade members shall have Operations", Rev 8E Enclosure Fire Brigade is verified to ensure that at least sufficient training and knowledge of nuclear A the Fire Brigade Leader and at least two Fire safety systems to understand the effects of fire OAP100.6 "Control Room Brigade members have sufficient knowledge of and fire suppressants on nuclear safety Conduct and Control of Shift nuclear safety systems to understand the performance Exception:

Sufficient training Activities", Rev 2E Attachment effects of fire and fire suppressants on nuclear and knowledge shall be permitted to be VIIA safety performance, unless the Exception is provided by an operations advisor dedicated to DC0780D-008 Industrial Fire used (Table S-2, Item 6)industrial fire brigade support criteria.

Brigade, Rev 0 (d)

  • The industrial fire brigade shall be CA EPP-013 "Fire Emergency", The Control Room notifies the Fire Brigade notified immediately upon verification of a fire. Rev 8 upon verification of a fire.DC0780D-009 Industrial Fire Brigade NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-18 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary (e) Each industrial fire brigade member shall C SAP-1 160 "Medical An annual physical examination is conducted pass an annual physical examination to Requirements for Special to assure each Fire Brigade member is determine that he or she can perform the Duties", Rev 9B Enclosure E qualified for respirator use. The annual strenuous activity required during manual fire- Fire Brigade Members physical examination is also conducted to fighting operations.

The physical examination assure the Fire Brigade Member can perform shall determine the ability of each member to the strenuous activity required during manual use respirator.

fire fighting operations (Table S-2, Item 6)3.4.2* Pre-Fire Plans: Current and detailed pre- C SAP-131 "Fire Protection Current, detailed Fire Pre-Plans are fire plans shall be available to the industrial fire Program", Rev 6D available to the Fire Brigade Leader and brigade for all areas in which a fire could Control Room personnel supporting the jeopardize the ability to meet the performance response to a fire event at the station (Table S-criteria described in Section 1.5. 2, Item 5)3.4.2.1* The plans shall detail the fire area C FPP031"Development and Fire Pre-Plans provide graphic and text configuration and fire hazards to be Control of Fire Protection representation of area configuration, area encountered in the fire area, along with any Preplans, Rev 3B hazards, FP Features and major nuclear safety nuclear safety components and fire protection components. (Table S-2, Item 5)systems and features that are present.3.4.2.2 Pre-fire plans shall be reviewed and updated C FPP031"Development and Station Modifications (e.g. Interface Review)as necessary.

Control of Fire Protection and Fire Drill Critique (e.g. Corrective Actions)Preplans, Rev 3B are mechanisms employed for Fire Pre-Plan MD-21 "Configuration usability, accuracy and improvements (Table Management", Rev 7 S-2, Item 5)SAP-131 "Fire Protection Program", Rev 6D SAP-133 "Design Control Program", Rev 14B EPP-13 "Fire Emergency", Rev 8 3.4.2.3* Pre-fire plans shall be available in the C SAP-131 "Fire Protection Current, detailed Fire Pre-Plans are control room and made available to the plant Program", Rev 6D available to the Fire Brigade Leader and industrial fire brigade. Control Room personnel supporting the response to a fire event at the station (Table S-2, Item 5)NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-19 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-19 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary 3.4.2.4* Pre-fire plans shall address coordination CA EPP-013 "Fire Emergency", Station fire response procedures and Fire with other plant groups during fire Rev 8 Brigade Leader Training discuss coordination emergencies.

TQP-606 "General Employee with other groups during fire emergencies.

Training Fire Protection Training" Rev 1A 3.4.3 Training and Drills Industrial fire brigade members and other C TQP-606 "General Employee Fire Brigade and other support personnel are plant personnel who would respond to a fire in Training Fire Protection Training" trained commensurate with their emergency conjunction with the brigade shall be provided Rev 1A responsibilities. (Table S-2, Item 6) (Table S-2, with training commensurate with their EPP-013 "Fire Emergency", Rev Item 7)emergency responsibilities.

8 EP-100 "Radiation Emergency Plan" Rev 59 (a) Plant Industrial Fire Brigade Training.

All of NRR Elements of the fire prevention program are the following requirements shall apply: described in the following subsections.

(1) Plant industrial fire brigade members CE TQP-606 "General Employee The station fire brigade has been evaluated shall receive training consistent with the Training Fire Protection Training" against the requirements of NFPA 600, requirements contained in NFPA 600, Rev 1A "Standard on Industrial Fire Brigades" (Table Standard on Industrial Fire Brigades, or NFPA DC0780D-009 "Industrial Fire S-2, Item 6)1500, Standard on Fire Department Brigade", Rev 0 Occupational Safety and Health Program, as appropriate.

(2) Industrial fire brigade members shall be C EPP-107 "Fire Brigade", Rev Quarterly training and practice is conducted given quarterly training and practice in fire OA in fire fighting, including radioactivity and health fighting, including radioactivity and health TQP-606 "General Employee physics considerations.

physics considerations, to ensure that each Training Fire Protection Training" member is thoroughly familiar with the steps to Rev 1A be taken in the event of a fire.(3) A written program shall detail the C TQP-606 "General Employee The fire brigade training program is industrial fire brigade training program. Training Fire Protection Training" documented in the station training procedure Rev 1A (4) Written records that include but are not CA Electronic Training Electronic Records are maintained for each limited to initial industrial fire brigade Qualification Program (Plateau)

Fire Brigade Member for Fire Brigade of NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-20 4 pI'- _'4W9:25AiR8 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary classroom and hands-on training, refresher TQP-409 "Implementation of training related activities including, but not training, special training schools attended, drill Training" Rev 0D limited to, classroom sessions, schools, drills attendance records, and leadership training for SAP-126, "Transmittal and and other related topics.industrial fire brigades shall be maintained for Maintenance of Records" Rev each industrial fire brigade member. 3A TQP-1004 "Training Documentation and Records", Rev 1A (b) Training for Non-Industrial Fire Brigade C EPP-013 "Fire Emergency", Fire Brigade support personnel are trained Personnel.

Plant personnel who respond with Rev 8 commensurate with their emergency the industrial fire brigade shall be trained as to responsibilities, potential hazards and their responsibilities, potential hazards to be interfacing with the Fire Brigade. (Table S-2, encountered, and interfacing with the industrial Item 7)fire brigade.(c)

  • Drills. All of the following requirements C EPP-107 "Fire Brigade", Rev Fire Brigade Drills are conducted on a shall apply. (1) Drills shall be conducted OA quarterly basis for each shift quarterly for each shift to test the response capability of the industrial fire brigade.(2) Industrial fire brigade drills shall be C EPP-1 07 "Fire Brigade", Rev Fire Brigade Drills are conducted with the developed to test and challenge industrial fire OA cited specific objectives to assess adequacy of brigade response, including brigade Fire Brigade response as a team during the performance as a team, proper use of drill scenario.equipment, effective use of pre-fire plans, and coordination with other groups. These drills shall evaluate the industrial fire brigade's abilities to react, respond, and demonstrate proper fire-fighting techniques to control and extinguish the fire and smoke conditions being simulated by the drill scenario.(3) Industrial fire brigade drills shall be C EPP-107 "Fire Brigade", Rev Guidance concerning the development of conducted in various plant areas, especially in OA Fire Brigade Drills, including important those areas identified to be essential to plant considerations (e.g. essential to plant operation and to contain significant fire operations, significant fire hazards) concerning hazards. drill scenario development are defined in NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-21 4r0'4=WC3eWffiG6 RC-11-0149 Attachment A NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary station procedures. (Table S-2, Item 7)(4) Drill records shall be maintained detailing C EPP-107 "Fire Brigade", Rev Station procedures define the maintenance the drill scenario, industrial fire brigade OA requirements for Fire Brigade Drill Records for member response, and ability of the industrial the Fire Brigade Members including the cited fire brigade to perform as a team. requirements. (Table S-2, Item 7)(5) A critique shall be held and documented C EPP-107 "Fire Brigade", Rev A critique is held and documented following after each drill. OA each Fire Brigade Drill.3.4.4 Fire-Fighting Equipment Protective clothing, respiratory protective C SAP-131 "Fire Protection Station procedures ensure that Fire Brigade equipment, radiation monitoring equipment, CE Program", Rev 6D personnel are provided with protective clothing personal dosimeters, and fire suppression DC0780D-001 "NFPA Code of and appropriate equipment.

equipment such as hoses, nozzles, fire Record", Enclosure C, Rev 0 extinguishers, and other needed equipment When acquired, the equipment conforms to shall be provided for the industrial fire brigade, applicable NFPA Standards.

These standards This equipment shall conform with the are identified in the identified evaluation (Table applicable NFPA standards.

S-2, Item 19)3.4.5 Off-Site Fire Department Interface 3.4.5.1 Mutual Aid Agreement Off-site fire authorities shall be offered a C EP-100, "Radiation Mutual Aid agreements have been plan for their interface during fires and related Emergency Plan" Letters of established (Letter Agreements) with offsite emergencies on site. Agreement (EP), Rev 59 organizations to respond to the station. A plan South Carolina Emergency for that interface during fire emergencies has Operation Plan, Annex 4, Feb been provided to off site fire authorities. (Table 2010 S-2, Item 7)3.4.5.2* Site-Specific Training Fire fighters from the off-site fire authorities C EP-100 "Radiation Emergency Offsite fire companies are offered site who are expected to respond to a fire at the Plan" Rev 59 specific training and are invited to a Fire plant shall be offered site-specific training and C-FP-19 "Indoctrination of Brigade Drill at least annually.

Attendance shall be invited to participate in a drill at least Offsite Fire Department" records are recorded in the station records annually.

EPP-107 "Fire Brigade", Rev OA system. (Table S-2, Item 7)NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-22 4rw4%C*61W9 RC-11-0149 mm---Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary 3.4,5.3* Security and Radiation Protection Plant security and radiation protection plans C EPP-013 "Fire Emergency", Fire, Radiological and Security Emergency shall address off-site fire authority response.

Rev 8 plans include roles and provisions for EP-100 "Radiation Emergency assistance to Off-Site fire authorities at the Plan" Rev 59 annual drills and offsite assistance is escorted SSP-1 14 "Security Force upon arrival.Responsibilities During Emergencies" Rev 14 3.4.6* Communications An effective emergency communications C EPP013 "Fire Emergency", An effective communication system has capability shall be provided for the industrial Rev 15A been provided to support fire fighting fire brigade. ECR71553 Fire Communication operations.

Primary communication is over the System station page system, while radio communication serve to support direct communication with the Fire Brigade Leader (Table S-2, Item 8)3.5.1 A fire protection water supply of adequate NRR Individual elements are addressed in the reliability, quantity, and duration shall be following sections provided by one of the two following methods.(a) Provide a fire protection water supply of NRR N/A A single water supply source via Lake not less than two separate 300,000-gal Monticello, provides the water supply to the (1,135,500-L) supplies, or (b) Fire Protection water distribution system. This option not selected.NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-23 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-23 41rw-49CW-ý"O RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary (b) Calculate the fire flow rate for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. C DC07810-015 "Intermediate The size of the Monticello Reservoir greatly This fire flow rate shall be based on 500 gpm Building 412' & 436' Sprinkler exceeds maximum fire flow demands including (1892.5 L/min) for manual hose streams plus System Hydraulic Analysis", rev automatic systems and manual hose streams the largest design demand of any sprinkler or 2 (Typical), for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The water supply system is fixed water spray system(s) in the power block VCSNS DBD "Fire Protection capable of delivering water at sufficient flow as determined in accordance with NFPA 13, System" (FS), Rev 2E Table 6.2- and pressure with the hydraulically demanding Standard for the Installation of Sprinkler 1 "Fire Water Demand" leg of the distribution system out of service.Systems, or NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection.

The fire water supply shall be capable of delivering this design demand with the hydraulically least demanding portion of fire main loop out of service.3.5.2* The tanks shall be interconnected such that NRR Water storage tanks are not used as the fire pumps can take suction from either or primary water supply source for the Fire both. A failure in one tank or its piping shall not Protection Water Distribution System allow both tanks to drain. The tanks shall be designed in accordance with NFPA 22, Standard for Water Tanks for Private Fire Protection.

Exception No. 1: Water storage tanks shall not be required when fire pumps are able to take suction from a large body of water (such as a lake), provided each fire pump has its own suction and both suctions and pumps are adequately separated.

Exception No. 2: Cooling tower basins shall be an acceptable water source for fire pumps when the volume is sufficient for both purposes and water quality is consistent with the demands of the fire service.NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-24 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-24 RC-1 1-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary 3.5.3* Fire pumps, designed and installed in C Drawing 302-231 Sht 1 "Fire The Diesel Engine Driven and Electric Motor accordance with NFPA 20, Standard for the CE Service Pumps", Rev 37 Driven Fire pumps are designed and installed Installation of Stationary Pumps for Fire DC0780D-003 "Automatic Fire to individually supply 100 percent of the Protection, shall be provided to ensure that Suppression", Rev 0 required flow rate and pressure.100 percent of the required flow rate and pressure are available assuming failure of the Fire pumps are designed and installed in largest pump or pump power source. accordance with NFPA 20. The fire pumps has been evaluated against the requirements of NFPA 20, "Standard for the Installation of Stationary Pumps for Fire Protection" 3.5,4 At least one diesel engine-driven fire pump C XPP0134B, Diesel Engine A single, diesel engine-driven fire pump is or two more seismic Category I Class IE Driven Fire Pump provided that is capable of supplying the electric motor-driven fire pumps connected to Drawing 302-231 Sht 1 "Fire required flow rate and pressure.redundant Class IE emergency power buses Service Pumps", Rev 37 capable of providing 100 percent of the required flow rate and pressure shall be provided.3.5.5 Each pump and its driver and controls shall C Drawing 126-001 "CW Pump The fire pumps are separated by three-hour be separated from the remaining fire pumps CE House Plans" Rev 7 rated barriers in the circulating water pump and from the rest of the plant by rated fire TR0780E-006 Fire Protection house.barriers.

Features:

FPEE Fire Pump Separation, Rev 0 The electric fire pump, driver and controller are Fire Protection Evaluation not separated from all other plant equipment by Report, Section 5.E.2(c) rated fire barriers.

This was previously described in our response to APCSB 9.5-1 Appendix A, Section E.2 Fire Protection Water Supply Systems, item (c) where we indicated"The diesel-driven fire pump is separated from the electric motor-driven fire pump by a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rated fire barrier in the circulating water intake screen and pumphouse" 3.5.6 Fire pumps shall be provided with automatic C SP-360 "1974 Fire Pumps" Each of the two fire pumps is coordinated to start and manual stop only. start on a drop in system pressure, and require manual action at the respective pump controller to stop the pump.NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-25 RC-11-0149 Attachment A NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary 3.5.7 Individual fire pump connections to the yard C Drawing 302-231 Sheets 1 Each fire pump is provided with separate fire main loop shall be provided and separated "Fire Service Pumps", Rev 37 connections to the yard fire main loop with with sectionalizing valves between Drawing 302-231 Sheets 2 "Fire sectionalizing valves between connections.

connections.

Service Hydrant and Loop" Rev 11 3.5.8 A method of automatic pressure C Drawing 302-231 Sheets 1 A jockey pump provides automatic pressure maintenance of the fire protection water "Fire Service Pumps", Rev 37 maintenance for the fire protection water system shall be provided independent of the Drawing 302-231 Sheets 2 "Fire system.fire pumps. Service Hydrant and Loop" Rev 11 3.5.9 Means shall be provided to immediately C Drawings 228- 044 Series The fire detection and control system in the notify the control room, or other suitable control room provides notification of fire pump constantly attended location, of operation of controller alarms, including "Pump Running".fire pumps.3.5.10 An underground yard fire main loop, C SP-124 "Yard Fire Protection The underground yard fire main loop was designed and installed in accordance with CE System", Rev 1 , designed and installed to supply manual and NFPA 24, "Standard for the Installation of Drawing 302-231 Sheets 1 "Fire automatic water based suppression systems, Private Fire Service Mains and Their Service Pumps", Rev 37 to meet system demands.Appurtenances", shall be installed to furnish Drawing 302-231 Sheets 2 "Fire anticipated water requirements.

Service Hydrant and Loop" Rev The fire protection water and distribution 11 system has been evaluated against the DC0780D-003 "Automatic Fire requirements of NFPA 24, "Standard for the Suppression", Rev 0 Installation of Private Fire Service Mains and Their Appurtenances" NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-26 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-26 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary 3.5.11 Means shall be provided to isolate portions C Drawing 302-231 Series System layout, including sectionalizing of the yard fire main loop for maintenance or valves, is provided to allow isolation of various repair without simultaneously shutting off the sections of the fire water system for supply to both fixed fire suppression systems maintenance or repair without adversely and fire hose stations provided for manual impacting primary and backup systems.backup. Sprinkler systems and manual hose station standpipes shall be connected to the plant fire protection water main so that a single active failure or a crack to the water supply piping to these systems can be isolated so as not to impair both the primary and backup fire suppression systems.3.5.12 Threads compatible with those used by local C FS DBD "Fire Protection Fire hose threads provided at the station are fire departments shall be provided on all System", Rev 2E Section 4.1.5.4 NHT standard, which are compatible with local hydrants, hose couplings, and standpipe NELPIA File No. NS-202, 1976 fire departments risers. SP-124 "Yard Fire Protection System" Section 1:03.5 Exception:

Fire departments shall be NRR Exception not applicable to Water Supply permitted to be provided with adapters that system design allow interconnection between plant equipment and the fire department equipment if adequate training and procedures are provided.NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-27 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-27 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary 3.5.13 Headers fed from each end shall be C SP-124 "Yard Fire Protection Each interior header has a separate permitted inside buildings to supply both CE System", connection with shutoff valve to the fire sprinkler and standpipe systems, provided SP- 337 "Pipe Line protection water distribution system, and steel piping and fittings meeting the Specifications for Conventional supplies both sprinkler and standpipe systems, requirements of ANSI B31.1, Code for Power Piping" Rev 13 Line Specl75X, which uses pipe and fittings meeting the Piping, are used for the headers (up to and 176X requirements of ANSI B31.1. Each sprinkler including the first valve) supplying the sprinkler Drawing 302-231 Series and standpipe connection is provided with an systems where such headers are part of the TR0780E-005 "Fire OS&Y valve for isolation purposes.seismically analyzed hose standpipe system. Suppression:

Seismic Where provided, such headers shall be Standpipes", Rev 0 The standpipe and water distribution piping is considered an extension of the yard main not a seismically analyzed system, which was system. Each sprinkler and standpipe system consistent with NRC Branch Technical Position shall be equipped with an outside screw and (BTP) APCSB 9.5-1 Appendix A, that was yoke (OS&Y) gate valve or other approved applicable to VCSNS (plants docketed prior to shutoff valve. July 1, 1976). Appendix A modified the requirements for hose standpipe systems and deleted this seismic design requirement.

3.5.14* All fire protection water supply and fire C STP-128.002 "FP Valve FP water distribution system valves are suppression system control valves shall be Lineup", Rev 18 locked or sealed, and periodically inspected for under a periodic inspection program and shall position.be supervised by one of the following methods.(a) Electrical supervision with audible and NRR Electric Supervision is not used to monitor visual signals in the main control room or other valve position.suitable constantly attended location.(b) Locking valves in their normal position.

C SAP- 0140 "Plant Key Where valves are provided locks, keys are Keys shall be made available only to Control", Rev 6A controlled through the Station Operations authorized personnel.

organization.(c) Sealing valves in their normal positions.

C STP-128.002 "FP Valve FP water distribution system valves are This option shall be utilized only where valves Lineup", Rev 18 normally sealed for valves in the Owner are located within fenced areas or under the Controlled Area.direct control of the owner/operator.

NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-28 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-28 Aýý RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary 3.5.15 Hydrants shall be installed approximately C Drawing E-303 Series Hydrants are typically spaced less than 400 every 250 ft (76 m) apart on the yard main CE "Underground Drawings for Fire feet intervals along the loop and at least 50 system. A hose house equipped with hose Service Piping" feet from the buildings.

Hose Houses are and combination nozzle and other auxiliary Drawing 302-231 Series "Fire located less than 1000 foot intervals along the equipment specified in NFPA 24, Standard for Service Hydrants and Loop" yard fire main.the Installation of Private Fire Service Mains TR0780E-005 "Fire and Their Appurtenances, shall be provided at Suppression:

Fire Hydrant The hose house and equipment has been intervals of not more than 1000 ft (305 m) Separation", Rev 0 evaluated against the requirements of NFPA along the yard main system. DC0780D-003 "Automatic Fire 24, "Standard for the Installation of Private Fire Suppression", Rev 0 Service Mains and Their Appurtenances" Exception:

Mobile means of providing hose NRR A mobile means to provided hose and and associated equipment, such as hose carts equipment is not relied upon, as an alternate to or trucks, shall be permitted in lieu of hose provide equipment to hose houses. Sufficient houses. Where provided, such mobile hose houses and equipment are located along equipment shall be equivalent to the the fire service main, to meet the requirements equipment supplied by three hose houses. of 3.5.15.3.5.16* The fire protection water supply system shall C Drawing 302-231 Series The fire protection water supply system is a be dedicated for fire protection use only. dedicated system for fire protection use.Complies with Exception No.1 below.Exception No. 1: Fire protection water C DC07810-030 "CB Preaction The water supply system is designed to supply systems shall be permitted to be used System Hydraulic Calculations", deliver sufficient flow and pressure for the to provide backup to nuclear safety systems, Rev 1 largest fire demand, in addition to supporting provided the fire protection water supply Emergency Diesel Generator cooling.systems are designed and maintained to deliver the combined fire and nuclear safety flow demands for the duration specified by the applicable analysis.Exception No. 2: Fire protection water NRR NA to the Fire Protection Water Supply storage can be provided by plant systems System design for VCSNS serving other functions, provided the storage has a dedicated capacity capable of providing the maximum fire protection demand for the specified duration as determined in this section.NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-29 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section DescriDtion DisDosition Reference Document Results Summary 3.6.1 For all power block buildings, Class III standpipe and hose systems shall be installed in accordance with NFPA 14, Standard for the Installation of Standpipe, Private Hydrant, and Hose Systems.CE DC0780D-004 "Manual Fire Suppression", Rev 0 TR0780E-005 "Fire Suppression:

Standpipe and Hose Stations" Rev 0 The hose standpipe system is designed and installed in accordance with NFPA 14. The standpipe and hose systems has been evaluated against the requirements of NFPA 14, "Standard for the Installation of Standpipe, Private Hydrant, and Hose Systems" The standpipe and water distribution piping is not a seismically analyzed system nor a Class III system, which was consistent with NRC Branch Technical Position (BTP) APCSB 9.5-1 Appendix A, that was applicable to VCSNS (plants docketed prior to July 1, 1976).Appendix A modified the requirements for hose standpipe systems and did not discuss class of service, and deleted the seismic design requirement.

3.6.2 A capability shall be provided to ensure an CA DC07810-036 "Nozzle The design of the system ensures an adequate water flow rate and nozzle pressure Pressure at Hose Reels", Rev 0 adequate flow rate and nozzle pressure for for all hose stations.

This capability includes concurrent operation of manual and fixed the provision of hose station pressure reducers suppressions systems.where necessary for the safety of plant industrial fire brigade members and off-site fire Pressure reducers have not been provided or department personnel.

are necessary.

Training of fire brigade members addresses high pressure nature of the system.NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-3D NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-30 on.---41!w4W9AFM"G RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary 3.6.3 The proper type of hose nozzle to be C Drawing DC07810-036 Nozzle The hose nozzles at the station are designed supplied to each power block area shall be Pressure at Hose Reels for electrical hazard service (UL Listed), and based on the area fire hazards. The usual are provided with shutoff capability, including combination spray/straight stream nozzle shall the ability to control water flow from full open to not be used in areas where the straight stream a full closed position.can cause unacceptable damage or present an electrical hazard to fire-fighting personnel.

Listed electrically safe fixed fog nozzles shall be provided at locations where high-voltage shock hazards exist. All hose nozzles shall have shutoff capability and be able to control water flow from full open to full closed.3.6.4 Provisions shall be made to supply water at C Drawing 302-231 Series The water supply and distribution system, least to standpipes and hose stations for CE TR0780E-005 "Fire supplies water to including fire hydrants, manual fire suppression in all areas containing Suppression:

Standpipe and standpipes and hose stations which are systems and components needed to perform Hose Stations" Rev 0 installed to support manual fire fighting the nuclear safety functions in the event of a operations in all areas containing systems or safe shutdown earthquake (SSE). components needed to perform nuclear safety functions.

The standpipe and water distribution piping is not a seismically analyzed system nor a Class III system, which was consistent with NRC Branch Technical Position (BTP) APCSB 9.5-1 Appendix A, that was applicable to VCSNS (plants docketed prior to July 1, 1976).Appendix A modified the requirements for hose standpipe systems and did not discuss class of service, and deleted the seismic design requirement.

NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-31 wm.---41rw4W9;rRdTrG8 RC-11-0149 Attachment A RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary Exception:

For existing plants that are not C EPP-027 "Hostile Action", Rev Isolation and restoration plans are available capable of meeting this requirement, 4B for beyond design basis events for the re-provisions to restore a water supply and establishment of fire service water distribution distribution system for manual fire-fighting system. THIS EXCEPTION OF THE purposes shall be made. This provisional REQUIREMENT IS NOT ENDORSED BY THE manual fire-fighting standpipe/hose station NRC.system shall be capable of providing manual fire-fighting protection to the various plant locations important to supporting and maintaining the nuclear safety function.

The provisions for establishing this provisional system shall be preplanned and be capable of being implemented in a timely manner following an SSE.3.6.5 Where the seismic required hose stations NRR There are no "seismic required" hose are cross-connected to essential seismic non- stations at VCSNS. Therefore degradation of fire protection water supply systems, the fire the supply as discussed in section 3.6.5 is not flow shall not degrade the essential water a present design consideration.

system requirement.

Where provided, fire extinguishers of the CE Station Fire Preplans, Fire extinguisher selection and layout are in appropriate number, size, and type shall be DC0780A-010 "Fire Extinguisher accordance with NFPA 10. The fire provided in accordance with NFPA 10, Layout", Rev 0 extinguishers have been evaluated against the Standard for Portable Fire Extinguishers.

DC0780D-005 "Manual requirements of NFPA10, "Standard for Extinguishers shall be permitted to be Suppression", Rev 0 Portable Fire Extinguishers" positioned outside of fire areas due to radiological conditions.

3.8.1 Fire Alarm NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-32 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-32 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary Alarm initiating devices shall be installed in C Drawings 228-044 Series Fire Alarm and Control System, which accordance with NFPA 72, National Fire Alarm CE Vendor Manual 1MS-94B-1334 includes the alarm initiating devices is Code. Alarm annunciation shall allow the "Simplex:

Fire Alarm 4100 designed as a proprietary alarm system such proprietary alarm system to transmit fire- Panel" that system alarms, supervisory signals and related alarms, supervisory signals, and Vendor Manual 1MS-94B-1335 trouble signals are transmitted to the Control trouble signals to the control room or other "Simplex:

Fire Alarm 2120 Room, to support timely response by station constantly attended location from which Panel" personnel.

required notifications and response can be EPP-013 "Fire Emergency", Rev initiated.

Personnel assigned to the 8 The Fire Detection and Control system has proprietary alarm station shall be permitted to DC0780D-005 "Fire Alarm and been evaluated against the requirements of have other duties. The following fire-related Detection Systems", Rev 0 NFPA 72 "National Fire Alarm Code" signals shall be transmitted:

(1) Actuation of any fire detection device C Vendor Manual 1 MS-94B- Actuation of any detection device is alarmed 1334 "Simplex:

Fire Alarm 4100 in the control room.Panel" Vendor Manual 1 MS-94B-1 335"Simplex:

Fire Alarm 2120 Panel" Drawings 228-044 Series "Fire Service Interconnection and Block Diagrams" SP-0928 "Fire Detection

&Control System", Rev 0 (2) Actuation of any fixed fire suppression C Vendor Manual 1MS-94B- Actuation of any fixed suppression is system 1334 "Simplex:

Fire Alarm 4100 alarmed in the control room.Panel" Vendor Manual 1MS-94B-1335"Simplex:

Fire Alarm 2120 Panel" Drawings 228-044 Series "Fire Service Interconnection and Block Diagrams" SP-0928 "Fire Detection

&Control System", Rev 0 NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-33 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-33 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary (3) Actuation of any manual fire alarm C Vendor Manual 1 MS-94B- Actuation of any manual fire alarm station is station 1334 "Simplex:

Fire Alarm 4100 alarmed in the control room.Panel" Vendor Manual 1 MS-94B-1 335"Simplex:

Fire Alarm 2120 Panel" Drawings 228-044 Series "Fire Service Interconnection and Block Diagrams" SP-0928 "Fire Detection

&Control System", Rev 0 (4) Starting of any fire pump C Vendor Manual 1 MS-94B- Starting of any fire pump is alarmed in the 1334 "Simplex:

Fire Alarm 4100 control room.Panel" Vendor Manual 1MS-94B-1335"Simplex:

Fire Alarm 2120 Panel" Drawings 228-044 Series "Fire Service Interconnection and Block Diagrams" SP-0928 "Fire Detection

&Control System", Rev 0 (5) Actuation of any fire protection C Vendor Manual 1 MS-94B- Actuation of any fire protection supervisory supervisory device 1334 "Simplex:

Fire Alarm 4100 device is alarmed in the control room.Panel" Vendor Manual 1MS-94B-1 335"Simplex:

Fire Alarm 2120 Panel" Drawings 228-044 Series "Fire Service Interconnection and Block Diagrams" SP-0928 "Fire Detection

&Control System", Rev 0 NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-34 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-34 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary (6) Indication of alarm system trouble C Vendor Manual 1MS-94B- Alarm system trouble condition is alarmed in condition 1334 "Simplex:

Fire Alarm 4100 the control room.Panel" Vendor Manual 1MS-94B-1335"Simplex:

Fire Alarm 2120 Panel" Drawings 228-044 Series "Fire Service Interconnection and Block Diagrams" SP-0928 "Fire Detection

&Control System", Rev 0 3.8.1.1 Means shall be provided to allow a person C EPP-013 "Fire Emergency", The plant page and station telephone observing a fire at any location in the plant to Rev 8 systems are the most common means for quickly and reliably communicate to the control reporting a fire to the control room, from any room or other suitable constantly attended location in the plant.location.3.8.1.2 Means shall be provided to promptly notify C EPP-013 "Fire Emergency", The Page Party System is used as the the following of any fire emergency in such a Rev 8 primary means to notify the General Site way as to allow them to determine an population and fire brigade members of the appropriate course of action: (1) General site appropriate course of action in the event of a population in all occupied areas; (2) Members fire emergency.

Communications with Off-Site of the industrial fire brigade and other groups agencies may be made by Land line telephone supporting fire emergency response; (3) Off- and radio system.site fire emergency response agencies.

Two independent means shall be available (e.g., telephone and radio) for notification of off-site emergency services.3.8.2 Detection NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-35 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-35 RC-11-0149 Attachment A RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary If automatic fire detection is required to meet CE SP-0928 "Fire Detection

& The fire alarm and detection system was the performance or deterministic requirements Control System", Rev 0 installed in accordance with NFPA 72, as of Chapter 4, then these devices shall be DC0780D-005 "Fire Alarm and described in NFPA 805, section 3.8.1.installed in accordance with NFPA 72, National Detection System", Rev 0 Fire Alarm Code, and its applicable Automatic fire detectors were designed and appendixes.

installed in accordance with NFPA 72E. The Fire Detection devices has been evaluated against the requirements of NFPA 72E"Standard for Automatic Fire Detectors" [Code of Record for original installation] (Table S-2, Item 10)3.9.1* If an automatic or manual water-based fire NRR Individual elements are addressed in the suppression system is required to meet the following sections performance or deterministic requirements of Chapter 4, then the system shall be installed in accordance with the appropriate NFPA standards including the following:

(1) NFPA 13, Standard for the Installation of CE DC0780D-003 Automatic Fire Sprinkler Systems are designed and Sprinkler Systems Suppression:

NFPA 13 installed in accordance with NFPA 13. The"Required" sprinkler systems have been evaluated against the requirements of NFPA 13, "Standard for the Installation of Sprinkler Systems" (2) NFPA 15, Standard for Water Spray NRR NA "Required" water spray systems that have Fixed Systems for Fire Protection not been identified as being credited for the NSCA or FirePRA Analysis.(3) NFPA 750, Standard on Water Mist Fire NRR NA VCSNS does not utilize Water Mist Fire Protection Systems Protection Systems.(4) NFPA 16, Standard for the Installation of NRR NA VCSNS does not utilize Foam Water Fire Foam-Water Sprinkler and Foam-Water Spray Protection Systems.Systems NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-36 A- a RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary 3.9.2 Each system shall be equipped with a water C Drawings 302-231 Series Each fire suppression system is equipped flow alarm. SP-0928 Fire Detection

& with a water flow alarm, which alarms to the Control System, Rev 0 Control Room.3.9.3 All alarms from fire suppression systems C Vendor Manual 1MS-94B- For water based systems, the fire alarm and shall communicate in the control room or other 1334 "Simplex:

Fire Alarm 4100 control system monitors and controls the suitable constantly attended.

Panel" transmission of water flow alarms to the Vendor Manual 1MS-94B-1335 Control Room."Simplex:

Fire Alarm 2120 Panel" Drawings 228-044 Series Fire Service Interconnection and Block Diagrams, SP-0928"Fire Detection

&Control System", Rev 0 3.9.4 Diesel-driven fire pumps shall be protected C Drawing 1MS-55-085-0026, The Diesel engine driven fire pump room is by automatic sprinklers. "Diesel Fire Pump Room protected by an automatic sprinkler system.Sprinkler" 3.9.5 Each system shall be equipped with an C Drawings 302-231 Series Each suppression system is equipped with OS&Y gate valve or other approved shutoff an OS&Y isolation valve.valve.3,9.6 All valves controlling water-based fire C See Section 3.5.14 Valves are normally sealed in the normal suppression systems required to meet the operating position.

See Section 3.5.14 performance or deterministic requirements of Chapter 4 shall be supervised as described in 3.5.14.3.10.1 If an automatic total flooding and local NRR Individual elements are addressed in the application gaseous fire suppression system is following sections required to meet the performance or deterministic requirements of Chapter 4, then the system shall be designed and installed in accordance with the following applicable NFPA codes: NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-37 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-37 41!N-4%9;F-AG*

RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary (1) NFPA 12, Standard on Carbon Dioxide CE DC0780D-003 Automatic Fire Low Pressure Carbon Dioxide Systems are Extinguishing Systems Suppression designed and installed in accordance with NFPA 12. The "Required" carbon dioxide system has been evaluated against the requirements of NFPA 12, "Standard on Carbon Dioxide Extinguishing Systems (2) NFPA 12A, Standard on Halon 1301 Fire NRR NA "Required" Halon 1301 systems have not Extinguishing Systems been identified as being credited for the NSCA or FirePRA Analysis.(3) NFPA 2001, Standard on Clean Agent NRR NA VCSNS does not utilize Clean Agent Fire Fire Extinguishing Systems Extinguishing Systems.3.10.2 Operation of gaseous fire suppression C SP-0928 "Fire Detection

& Gaseous fire suppression systems alarms systems shall annunciate and alarm in the Control System" are monitored by the fire alarm and control control room or other constantly attended Vendor Manual 1MS-94B-1334 system, which annunciates this condition to the location identified. "Simplex:

Fire Alarm 4100 Control Room.Panel" Vendor Manual 1MS-94B-1 335"Simplex:

Fire Alarm 2120 Panel" Drawings 228-044 Series Fire Service Interconnection and Block Diagrams, 3.10.3 Ventilation system design shall take into C VCSNS DBD "Fire Protection Functional testing addressed loss of agent account prevention from over pressurization System" (FS), Rev 2E via discharge testing. No indications of over during agent injection, adequate sealing to Section 3.4.4 pressurization were evident in system files.prevent loss of agent, and confinement of Chemtron Ltr dated, 3/8/82, The C02 system(s) are not installed in radioactive contaminants.

Field Test Report Job# FL radiological areas of the plant.22425-3 3.10.4* In any area required to be protected by both NRR NA There are no installed backup gaseous fire primary and backup gaseous fire suppression suppression systems utilized at VCSNS.systems, a single active failure or a crack in any pipe in the fire suppression system shall not impair both the primary and backup fire suppression capability.

NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-38 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary 3.10.5 Provisions for locally disarming automatic C SAP-201 "Equipment Tagging The disarmed status of the C02 system is gaseous suppression systems shall be and Lockout -Tag Out", Rev 11 C annunciated in the Control Room to assure that secured and under strict administrative control. FS DBD "Fire Protection", Rev the system is returned to its operating (armed)2E status when the area is no longer occupied.Drawing 302-232 "P&ID: FS Positive system status control is maintained by Halon and Low Pressure C02", Control Room personnel for this system.Rev 5 3.10.6* Total flooding carbon dioxide systems shall C Field Walkdown C02 systems in CB06, CB07 & CB09 are not be used in normally occupied areas. not located in areas normally inhabited by station personnel, do not have permanent work stations, are not normally utilized as an occupied area, and are not a normal personnel pass through area.3.10.7 Automatic total flooding carbon dioxide C SP-117 "Plant Fire Protection Pre-discharge alarms, time delays and systems shall be equipped with an audible pre- System" odorizers are provided personnel protection for discharge alarm and discharge delay sufficient Drawing 302-233 "Halon and total flooding CO 2 protected areas.to permit egress of personnel.

The carbon Low Pressure C02", Rev 5 dioxide system shall be provided with an odorizer.3.10.8 Positive mechanical means shall be C SAP-201 "Equipment Tagging Isolation valves are provided as a provided to lock out total flooding carbon and Lockout -Tag Out", Rev 1 iC mechanical means to isolate the C02 dioxide systems during work in the protected FS DBD "Fire Protection", Rev system(s).

space. 2E Drawing 302-232 "P&ID: FS Halon and Low Pressure C02", Rev 5 3.10.9 The possibility of secondary thermal shock CE Field Walkdown The placement of discharge nozzles to (cooling) damage shall be considered during TR0780E-005 "Fire sensitive electrical equipment was considered the design of any gaseous fire suppression Suppression:

C02 Thermal during design. Nozzle positions were reviewed system, but particularly with carbon dioxide. Shock", Rev 0 for possible impacts.NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-39 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-39 Aýý 4 RC-1 1-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary 3.10.10 Particular attention shall be given to C SP-117 "Plant Fire Protection An odorless gas, which is non conductive, corrosive characteristics of agent System" carbon dioxide [C02] is widely distributed in decomposition products on safety systems. nature and is a minor component of air. It is highly soluble in water and oil. The presence of C02 in water can create a weak acid, however the agent is not normally present in the protected area. The C02 system should extinguish a fire prior to application of water from fire hoses. However, post fire cleanup, depending on extent of fire damage shall consider potential impact of decomposition products on safety systems.This section shall be used to determine the NRR Individual elements are addressed in the design and installation requirements for following sections passive protection features.

Passive fire protection features include wall, ceiling, and floor assemblies, fire doors, fire dampers, and through fire barrier penetration seals. Passive fire protection features also include electrical raceway fire barrier systems (ERFBS) that are provided to protect cables and electrical components and equipment from the effects of fire.3.11.1 Building Separation Each major building within the power block C Drawings E023 Series Major buildings within the power block are shall be separated from the others by barriers separated by 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> construction.

having a designated fire resistance rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or by open space of at least 50 ft (15.2 m) or space that meets the requirements of NFPA 80A, "Recommended Practice for Protection of Buildings from Exterior Fire Exposures." NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-40 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-40 w=-__'41!04W9AF-4&G*

RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary Exception:

Where a performance-based NRR analysis determines the adequacy of building separation, the requirements of 3.11.1 shall not apply.3.11.2 Fire Barriers Fire barriers required by Chapter 4 shall CE TR0780E-006 "Fire Protection The fire barriers have been designed or include a specific fire resistance rating. Fire Features:

Fire Barriers", Rev 0 evaluated to satisfy the performance barriers shall be designed and installed to requirements of Chapter 4. Where applicable, meet the specific fire resistance rating using references to standard qualification tests have assemblies qualified by fire tests. The been included in evaluations.

qualification fire tests shall be in accordance with NFPA 251, Standard Methods of Tests of Equivalency provided to describe overall fire Fire Endurance of Building Construction and barriers construction details has been Materials, or ASTM E 119, Standard Test documented in selected cases to support Methods for Fire Tests of Building equivalent performance to tested Construction and Materials.

configurations. (Table S-2, Item 11)3.11.3* Fire Barrier Penetrations 3.11.3 Penetrations in fire barriers shall be CE DC0780D-007 Passive Fire Fire Doors and Dampers are evaluated in provided with listed fire-rated door assemblies Protection Features, Rev 0 openings to ensure the rating of the penetrated or listed rated fire dampers having a fire fire barrier. (Table S-2, Item 11)resistance rating consistent with the designated fire resistance rating of the barrier as determined by the performance requirements established by Chapter 4. (See 3.11.4 for penetration seals for through penetration fire stops.) Passive fire protection devices such as doors and dampers shall conform with the following NFPA standards, as applicable:

NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-41 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-41 40- 9 ° RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary (1) NFPA 80, Standard for Fire Doors and CE DC0780D-007 Passive Fire The fire door design has been evaluated Fire Windows CNRC Protection Features, Rev 0 against the requirements of NFPA 80 TR0787E-006 "Fire Protection "Standard for Fire Doors and Windows".Features:

Specialty Fire Doors", Rev 0 Prior NRC approval of specialty doors (e.g.pressure and bullet resistant) were previously found to be acceptable and has been summarized in the defined evaluation (2) NFPA 90A, Standard for the Installation CE DC0780D-007 Passive Fire The fire damper design has been evaluated of Air-Conditioning and Ventilating Systems CNRC Protection Features, Rev 0 against the requirements of NFPA 90A TR0787E-006 "Fire Protection "Standard for the Installation of Air Features:

Fire Dampers, Back to Conditioning and Ventilating Systems".Back", Rev 0 Prior NRC approval of unique damper configurations were previously found to be acceptable and has been summarized in the defined evaluation (3) NFPA 101, Life Safety Code NRR NFPA101 is exempted from the scope of the NRC review per 10 CFR 50.48 C.2 (i)regarding Life Safety. The requirements related to fire doors /fire dampers are addressed in the NFPA 80 and NFPA 90A code compliance reviews.Exception:

Where fire area boundaries are C TR0780E-001 "NSCA Fire rated boundaries, when identified have not wall-to-wall, floor-to-ceiling boundaries with Separation", Rev 0 openings which are protected as described in all penetrations sealed to the fire rating TR07870-001 "Fire Rated 3.11.3. In cases where boundaries were required of the boundaries, a performance-Seals", Rev 0 credited where the boundary was not wall to based analysis shall be required to assess the DC0780C Series Calculations" wall or floor to ceiling, with all opening adequacy of fire barrier forming the fire Multicompartment Analysis" protected, a performance based analysis has boundary to determine if the barrier will been performed to assess the adequacy of the withstand the fire effects of the hazards in the fire barrier forming the fire boundary area. Openings in fire barriers shall be permitted to be protected by other means as acceptable to the AHJ.NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-42 nw-_ RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary 3.11.4* Through Penetration Fire Stops Through penetration fire stops for C TR07870-002 "Penetration Penetration seals through rated fire barriers penetrations such as pipes, conduits, bus CE Seals Engineering Evaluations", are qualified by test or evaluated to maintain ducts, cables, wires, pneumatic tubes and Rev 0 the rating of the penetrated fire barrier. (Table ducts, and similar building service equipment S-2, Item 12)that pass through fire barriers shall be protected as follows: (a) The annular space between the C TR07870-002 "Penetration Penetration seals through rated fire barriers penetrating item and the through opening in CE Seals Engineering Evaluations", are qualified by test or evaluated to maintain the fire barrier shall be filled with a qualified Rev 0 the rating of the penetrated fire barrier. (Table fire-resistive penetration seal assembly S-2, Item 11)capable of maintaining the fire resistance of the fire barrier. The assembly shall be qualified by tests in accordance with a fire test protocol acceptable to the AHJ or be protected by a listed fire-rated device for the specified fire-resistive period.(b) Conduits shall be provided with an C Drawing 201-240, Sheet 2 Internal conduit penetration seals are closed internal fire seal that has an equivalent fire- "Fire, Pressure and Radiation with internal fire seals to maintain the rating of resistive rating to that of the fire barrier through Barrier Details", Rev 11 the barrier.opening fire stop and shall be permitted to be installed on either side of the barrier in a location that is as close to the barrier as possible.Exception:

Openings inside conduit 4 in. C Drawing 201-240, Sheet 2 Internal conduit penetration seals are closed (10.2 cm) or less in diameter shall be sealed at "Fire, Pressure and Radiation with internal fire seals to maintain the rating of the fire barrier with a fire-rated internal seal Barrier Details", Rev 11 the barrier as described in the Exception.

unless the conduit extends greater than 5 ft (1.5 m) on each side of the fire barrier. In this case the conduit opening shall be provided with noncombustible material to prevent the passage of smoke and hot gases. The fill depth of the material packed to a depth of 2 in.(5.1 cm) shall constitute an acceptable smoke NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-43

.--* RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary and hot gas seal in this application.

3.11.5* Electrical Raceway Fire Barrier Systems (ERFBS)ERFBS required by Chapter 4 shall be C TR07870-001 "Kaowool Triple "Required" ERFBS are installed to provide 1-capable of resisting the fire effects of the CE Wrap Raceway Fire Barrier Test hour or 3-hour fire barrier rating. Qualification hazards in the area. ERFBS shall be tested in for Conduits and Cable Tray", testing has been performed in accordance with accordance with and shall meet the Rev 0 Generic Letter 86-10, Supplement 1 or acceptance criteria of NRC Generic Letter 86- TR07870-017 "Evaluation of equivalent performance testing to ensure the 10, Supplement 1, "Fire Endurance Test Interam E-54A Fire Wrap" Rev 0 protected raceways are free of fire damage.Acceptance Criteria for Fire Barrier Systems DC07870-003 "Kaowool Thermal (Table S-2, Item 11)Used to Separate Safe Shutdown Trains Mass Comparison of Test vs.Within the Same Fire Area." The ERFBS Plant" Rev 0 A summary of ERFBS testing and systems needs to adequately address the design TR0780E-006 "Fire Protection employed at the station has been included in requirements and limitations of supports and Features:

ERFBS", Rev 0 the engineering evaluation.

intervening items and their impact on the fire barrier system rating. The fire barrier system's ability to maintain the required nuclear safety circuits free of fire damage for a specific thermal exposure, barrier design, raceway size and type, cable size, fill, and type shall be demonstrated.

Exception No. 1: When the temperatures C TR07870-001 "KaowoolTriple Guidance concerning cable functionality may inside the fire barrier system exceed the Wrap Raceway Fire Barrier Test be used, as appropriate, in the evaluation of maximum temperature allowed by the for Conduits and Cable Tray", performance of ERFBS, via electrical testing acceptance criteria of Generic Letter 86-10, Rev 0 requirements of Generic Letter 86-10,"Fire Endurance Acceptance Test Criteria for TR07870-017 "Evaluation of Supplement 1, Attachment 1.Fire Barrier Systems Used to Separate Interam E-54A Fire Wrap" Rev 0 Redundant Safe Shutdown Training Within the DC07870-003 "Kaowool Thermal Same Fire Area," Supplement 1, functionality Mass Comparison of Test vs.of the cable at these elevated temperatures Plant" Rev 0 shall be demonstrated.

Qualification demonstration of these cables shall be performed in accordance with the electrical testing requirements of Generic Letter 86-10, Supplement 1, Attachment 1, "Attachment NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-44 RC-11-0149 Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Section Section Description Disposition Reference Document Results Summary Methods for Demonstrating Functionality of Cables Protected by Raceway Fire Barrier Systems During and After Fire Endurance Test Exposure." Exception No. 2: ERFBS systems NRR ERFBS systems have been evaluated after employed prior to the issuance of Generic the issuance of Generic Letter 86-10, Letter 86-10, Supplement 1, are acceptable Supplement 1.providing that the system successfully met the limiting end point temperature requirements as specified by the AHJ at the time of acceptance.

NEI 04-02 Table B-I Transition of Fundamental Fire Protection Program & Design Elements Page A-45 NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Page A-45 A Jdýt*ý* RC-11-0149 Attachment B B. NEI 04-02 Table B-2 -Nuclear Safety Capability Assessement

-Methodology Review 40 Pages Attached NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-1 NFPA 805 Section 2.4.2 Nuclear Safety Capability Assessment Attachment B 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 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 occasion of nuclear safety equipment and cables (4) Assessment of the ability to achieve the nuclear safety performance criteria given in each fire area.Steps 1 through 4 shall be performed to determined equipment and cables that shall be evaluated using either the deterministic or performance-based method in Chapter 4.Other performance-based or risk-informed methods acceptable to authority having jurisdiction (AHJ) shall be permitted.

NEI 00-01 Section 3. Deterministic Methodology NEI 00-01 Section Description NFPA 805 Alignment NFPA 805 Reference This section discusses a generic deterministic methodology and criteria that licensees Introductory section, can use to perform a post-fire safe shutdown analysis to address regulatory alignment identified in requirements.

subsections NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-2 a A seýtdýAttachment B NFPA 805 Section 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire event shall be developed.

The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation of result in the maloperation of those components needed to meet the nuclear the safety criteria shall be included.

Availability and reliability of equipment selected shall be evaluated. (See Appendix B for acceptable methods used to identify equipment)

NEI 00-01 Section NEI 00-01 Section Description NFPA 805 Alignment NFPA 805 Reference 3.1 [A, Intro] Safe Shutdown Systems and Path Development 3.1 [B, Goals] Safe Shutdown Systems and Path Development 3.1 [C, Spurious Operations]

Safe Shutdown Systems and Path Development 3.1.1 Criteria/Assumptions This section discusses the identification of systems available and necessary to perform the required safe shutdown functions.

It also provides information on the process for combining these systems into safe shutdown paths. Appendix R Section IIl.G.1.a requires that the capability to achieve and maintain hot shutdown be free of fire damage. It is expected that the term "free of fire damage" will be further clarified in a forthcoming Regulatory Issue Summary. Appendix R Section III.G.1.b requires that repairs to systems and equipment necessary to achieve and maintain cold shutdown be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. It is the intent of the NRC that requirements related to the use of manual operator actions will be addressed in a forthcoming rulemaking.

The goal of post-fire safe shutdown is to assure that a one train of shutdown systems, structures, and components remains free of fire damage for a single fire in any single plant fire area. This goal is accomplished by determining those functions important to achieve and maintain hot shutdown.

Safe shutdown systems are selected so that the capability to perform these required functions is a part of each safe shutdown path.The functions important to post-fire safe shutdown generally include, but are not limited to the following:

-Reactivity control; -Pressure control systems; -Inventory control systems; -Decay heat removal systems; -Process monitoring;

-Support systems;-

Electrical systems; -Cooling systems. These functions are of importance because they have a direct bearing on the safe shutdown goal of being able to achieve and maintain hot shutdown which ensures the integrity of the fuel, the reactor pressure vessel, and the primary containment.

If these functions are preserved, then the plant will be safe because the fuel, the reactor and the primary containment will not be damaged. By assuring that this equipment is not damaged and remains functional, the protection of the health and safety of the public is assured.In addition to the above listed functions, Generic Letter 81-12 specifies consideration of associated circuits with the potential for spurious equipment operation and/or loss of power source, and the common enclosure failures.

Spurious operations/actuations can affect the accomplishment of the post-fire safe shutdown functions listed above.Typical examples of the effects of the spurious operations of concern are the following: -A loss of reactor pressure vessel/reactor coolant inventory in excess of the safe shutdown makeup capability; -A flow loss or blockage in the inventory makeup or decay heat removal systems being used for the required safe shutdown path.Spurious operations are of concern because they have the potential to directly affect the ability to achieve and maintain hot shutdown, which could affect the fuel and cause damage to the reactor pressure vessel or the primary containment.

Common power source and common enclosure concerns could also affect these and must be addressed.

The following criteria and assumptions may be considered when identifying systems available and necessary to perform the required safe shutdown functions and combining these systems into safe shutdown paths.Aligns See Nuclear Safety Equipment Technical Report TR08620-015 Aligns See Nuclear Safety Equipment Technical Report TR08620-015 Aligns See Nuclear Safety Equipment Technical Report TR08620-015 Aligns See Nuclear Safety Equipment Technical Report TR08620-015 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-3 Attachment B NFPA 805 Section 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection 3.1.1.01 [ BWR Safe Shutdown Paths]3.1.1.02 [BWR Safety Relief Valves/low Pressure Systems]3.1.1.03 [PWR, Pressurizer Heaters]3.1.1.04 [Alternative Shutdown Capability]

3.1.1.05 [Initial Conditions]

3.1.1.06 [Other Events in Conjunction with Fire]3.1.1.07 [Availability of Offsite Power][BWR] GE Report GE-NE-T43-00002-00-01-RO1 entitled "Original Safe Shutdown Paths For The BWR" addresses the systems and equipment originally designed into the GE boiling water reactors (BWRs) in the 1960s and 1970s, that can be used to achieve and maintain safe shutdown per Section II.G.1 of 10CFR 50, Appendix R.Any of the shutdown paths (methods) described in this report are considered to be acceptable methods for achieving redundant safe shutdown.[BWR] GE Report GE-NE-T43-00002-00-03-R01 provides a discussion on the BWR Owners' Group (BWROG) position regarding the use of Safety Relief Valves (SRVs)and low pressure systems (LPCI/CS) for safe shutdown.

The BWROG position is that the use of SRVs and low pressure systems is an acceptable methodology for achieving redundant safe shutdown in accordance with the requirements of 10CFR50 Appendix R Sections Ill.G.1 and III.G.2. The NRC has accepted the BWROG position and issued anSER dated Dec. 12, 2000.[PWR] Generic Letter 86-10, Enclosure 2, Section 5.3.5 specifies that hot shutdown can be maintained without the use of pressurizer heaters (i.e., pressure control is provided by controlling the makeup/charging pumps). Hot shutdown conditions can be maintained via natural circulation of the RCS through the steam generators.

The cooldown rate must be controlled to prevent the formation of a bubble in the reactor head. Therefore, feedwater (either auxiliary or emergency) flow rates as well as steam release must be controlled.

The classification of shutdown capability as alternative shutdown is made independent of the selection of systems used for shutdown.

Alternative shutdown capability is determined based on an inability to assure the availability of a redundant safe shutdown path. Compliance to the separation requirements of Sections II.G.1 and II.G.2 may be supplemented by the use of manual actions to the extent allowed by the regulations and the licensing basis of the plant, repairs (cold shutdown only), exemptions, deviations, GL 86-10 fire hazards analyses or fire protection design change evaluations, as appropriate.

These may also be used in conjunction with alternative shutdown capability.

At the onset of the postulated fire, all safe shutdown systems (including applicable redundant trains) are assumed operable and available for post-fire safe shutdown.Systems are assumed to be operational with no repairs, maintenance, testing, Limiting Conditions for Operation, etc. in progress.

The units are assumed to be operating at full power under normal conditions and normallineups.

No Final Safety Analysis Report accidents or other design basis events (e.g. loss of coolant accident, earthquake), single failures or non-fire induced transients need be considered in conjunction with the fire.For the case of redundant shutdown, offsite power may be credited if demonstrated to be free of fire damage. Offsite power should be assumed to remain available for those cases where its availability may adversely impact safety (i.e., reliance cannot be placed on fire causing a loss of offsite power if the consequences of offsite power availability are more severe than its presumed loss). No credit should be taken for a fire causing a loss of offsite power. For areas where train separation cannot be achieved and altemative shutdown capability is necessary, shutdown must be demonstrated both where offsite power is available and where offsite power is not available for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.Not Applicable Not Applicable Aligns See Nuclear Safety Equipment Technical Report TR08620-015 Aligns See Nuclear Safety Capability Assessment Technical Report TR08620-312 Aligns See Nuclear Safety Capability Assessment Technical Report TR08620-312 See Nuclear Safety Capability Assessment Technical Report TR08620-312 See Nuclear Safety Capability Assessment Technical Report TR08620-312 Aligns Aligns NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-4 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-4

' -4 Attachment B NFPA 805 Section 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection 3.1.1.08 [Safety-Related Post-fire safe shutdown systems and components are not required to be safety-related.

Aligns Equipment]

3.1.1.09 [72-hour Coping Period]3.1.1.10[Manual/Automatic Initiation of Systems]3.1.1.11 [Multiple Affected Units]The post-fire safe shutdown analysis assumes a 72-hour coping period starting with a reactor scram/trip.

Fire-induced impacts that provide no adverse consequences to hot shutdown within this 72-hour period need not be included in the post-fire safe shutdown analysis.

At least one train can be repaired or made operable within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> using onsite capability to achieve cold shutdown.Aligns See Nuclear Safety Equipment Technical Report TR08620-015 See Nuclear Safety Capability Assessment Technical Report TR08620-312 See Nuclear Safety Capability Assessment Technical Report TR08620-312 Manual initiation from the main control room or emergency control stations of systems Aligns required to achieve and maintain safe shutdown is acceptable where permitted by current regulations or approved by NRC; automatic initiation of systems selected for safe shutdown is not required but may be included as an option.Where a single fire can impact more than one unit of a multi-unit plant, the ability to achieve and maintain safe shutdown for each affected unit must be demonstrated.

Not Applicable NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-5

~iG., ~Attachment B NFPA 805 Section 2.4.2.1 Appendix B.1 Nuclear Safety Assessment The primary purpose of the nuclear safety assessment is to demonstrate that given cable and equipment damage due to a fire postulated in any fire area, sufficient equipment remains available to achieve the following nuclear safety performance criteria (see Section 1.5): (1) Reactivity control (2) Inventory and pressure control (3) Decay heat removal (4) Vital auxiliaries (5) Process monitoring The purpose of this appendix is to identify attributes that should be considered when demonstrating this capability.

Other risk informed-performance-based methods acceptable to the AHJ are permitted.

NEI 00-01 Section NEI 00-01 Section Description

3.1.2 Shutdown

Functions 3.1.2.1 Reactivity Control 3.1.2.2 Pressure Control Systems The following discussion on each of these shutdown functions provides guidance for selecting the systems and equipment required for safe shutdown.

For additional information on BWR system selection, refer to GE Report GENE- T43-00002-00 R01 entitled "Original Safe Shutdown Paths for the BWR."[BWR] Control Rod Drive System. The safe shutdown performance and design requirements for the reactivity control function can be met without automatic scram/trip capability.

Manual scram/reactor trip is credited.

The post-fire safe shutdown analysis must only provide the capability to manually scram/trip the reactor.[PWR]

Makeup/Charging There must be a method for ensuring that adequate shutdown margin is maintained by ensuring borated water is utilized for RCS makeup/charging.

The systems discussed in this section are examples of systems that can be used for pressure control. This does not restrict the use of other systems for this purpose.[BWR] Safety Relief Valves (SRVs)The SRVs are opened to maintain hot shutdown conditions or to depressurize the vessel to allow injection using low pressure systems.These are operated manually.

Automatic initiation of the Automatic Depressurization System is not a required function.[PWR]

Makeup/ChargingRCS pressure is controlled by controlling the rate of charging/makeup to the RCS. Although utilization of the pressurizer heaters and/or auxiliary spray reduces operator burden, neither component is required to provide adequate pressure control. Pressure reductions are made by allowing the RCS to cool/shrink, thus reducing pressurizer level/pressure.

Pressure increases are made by initiating charging/makeup to maintain pressurizer level/pressure.

Manual control of the related pumps is acceptable.

NFPA 805 Alignment Introductory section, alignment identified in subsections NFPA 805 Reference Aligns See Nuclear Safety Equipment Technical Report TR08620-015 Aligns See Nuclear Safety Equipment Technical Report TR08620-015 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-6 Attachment B NFPA 805 Section 2.4.2.1 Appendix B-1 Nuclear Safety Assessment 3.1.2.3 Inventory Control Systems 3.1.2.4 Decay Heat Removal Systems 3.1.2.5 Process Monitoring Systems selected for the inventory control function should be capable of maintaining level to achieve and maintain hot shutdown.

Typically, the same components providing inventory control are capable of providing pressure control. Manual initiation of these systems is acceptable.

Automatic initiation functions are not required.

[BWR]Systems selected for the inventory control function should be capable of supplying sufficient reactor coolant to achieve and maintain hot shutdown.

Manual initiation of these systems is acceptable.

Automatic initiation functions are not required.[PWR]:

Systems selected for the inventory control function should be capable of maintaining level to achieve and maintain hot shutdown.

Typically, the same components providing inventory control are capable of providing pressure control. Manual initiation of these systems is acceptable.

Automatic initiation functions are not required.[BWR] Systems selected for the decay heat removal function(s) should be capable of:- Removing sufficient decay heat from primary containment, to prevent containment over-pressurization and failure. -Satisfying the net positive suction head requirements of any safe shutdown systems taking suction from the containment (suppression pool). -Removing sufficient decay heat from the reactor to achieve cold shutdown.[PWR]

Systems selected for the decay heat removal function(s) should be capable of:- Removing sufficient decay heat from the reactor to reach hot shutdown conditions.

Typically, this entails utilizing natural circulation in lieu of forced circulation via the reactor coolant pumps and controlling steam release via the Atmospheric Dump valves.- Removing sufficient decay heat from the reactor to reach cold shutdown conditions.-

This does not restrict the use of other systems.The process monitoring function is provided for all safe shutdown paths. IN 84-09, Attachment 1,Section IX "Lessons Learned from NRC Inspections of Fire Protection Safe Shutdown Systems (10CFR50 Appendix R)" provides guidance on the instrumentation acceptable to and preferred by the NRC for meeting the process monitoring function.

This instrumentation is that which monitors the process variables necessary to perform and control the functions specified in Appendix R Section II.L.1.Such instrumentation must be demonstrated to remain unaffected by the fire. The IN 84-09 list of process monitoring is applied to alternative shutdown (III.G.3).

IN 84-09 did not identify specific instruments for process monitoring to be applied to redundant shutdown (llI.G.land III.G.2).

In general, process monitoring instruments similar to those listed below are needed to successfully use existing operating procedures (includingAbnormal Operating Procedures).

[BWR] -Reactor coolant level and pressure -Suppression pool level and temperature

-Emergency or isolation condenser level -Diagnostic instrumentation for safe shutdown systems -Level indication for tanks needed for safe shutdown[PWR] -Reactor coolant temperature (hot leg / cold leg) -Pressurizer pressure and level -Neutron flux monitoring (source range) -Level indication for tanks needed for safe shutdown -Steam generator level and pressure -Diagnostic instrumentation for safe shutdown systems -The specific instruments required may be based on operator preference, safe shutdown procedural guidance strategy (symptomatic vs.prescriptive), and systems and paths selected for safe shutdown.Aligns See Nuclear Safety Equipment Technical Report TR08620-015 Aligns See Nuclear Safety Equipment Technical Report TR08620-015 Aligns See Nuclear Safety Equipment Technical Report TR08620-015 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-7 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-7

.ýta Attachment B NFPA 805 Section 2.4.2.1 Appendix B.1 Nuclear Safety Assessment 3.1.2.6 Support Systems 3.1.2.6.1

[Electrical Systems]3.1.2.6.2

[A, HVAC Systems]3.1.2.6.2

[B, Cooling Systems]AC Distribution System Power for the Appendix R safe shutdown equipment is typically provided by a medium voltage system such as 4.16 KV Class 1E busses either directly from the busses or through step down transformers/load centers/distribution panels for 600, 480 or 120 VAC loads. For redundant safe shutdown performed in accordance with the requirements of Appendix R Section III.G.1 and 2, power may be supplied from either offsite power sources or the emergency diesel generator depending on which has been demonstrated to be free of fire damage. No credit should be taken for a fire causing a loss of offsite power. Refer to Section 3.1.1.7. DC Distribution System Typically, the 125VDC distribution system supplies DC control power to various 125VDC control panels including switchgear breaker controls.

The 125 VDC distribution panels may also supply power to the 120VAC distribution panels via static inverters.

These distribution panels typically supply power for instrumentation necessary to complete the process monitoring functions.

For fire events that result in an interruption of power to the AC electrical bus, the station batteries are necessary to supply any required control power during the interim time period required for the diesel generators to become operational.

Once the diesels are operational, the 125 VDC distribution system can be powered from the diesels through the battery chargers.

The DC control centers may also supply power to various small horsepower Appendix R safe shutdown system valves and pumps. If the DC system is relied upon to support safe shutdown without battery chargers being available, it must be verified that sufficient battery capacity exists to support the necessary loads for sufficient time (either until power is restored, or the loads are no longer required to operate).HVAC Systems -HVAC Systems may be required to assure that safe shutdown equipment remains within its operating temperature range, as specified in manufacturer's literature or demonstrated by suitable test methods, and to assure protection for plant operations staff from the effects of fire (smoke, heat, toxic gases, and gaseous fire suppression agents). HVAC systems may be required to support safe shutdown system operation, based on plant-specific configurations.

Typical uses include: -Main control room, cable spreading room, relay room; -ECCS pump compartments; -Diesel generator rooms; -Switchgear rooms. Plant-specific evaluations are necessary to determine which HVAC systems are essential to safe shutdown equipment operation.

Various cooling water systems may be required to support safe shutdown system operation, based on plant-specific considerations.

Typical uses include:

  • Diesel generator cooling;
  • RHR/SDC/DH Heat Exchanger cooling water, -Safe shutdown pump cooling (seal coolers, oil coolers);
  • HVAC system cooling water Aligns See Nuclear Safety Equipment Technical Report TR08620-015 Aligns See Nuclear Safety Equipment Technical Report TR08620-015 Aligns See Nuclear Safety Equipment Technical Report TR08620-015 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

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.At-to~Attachment B NFPA 805 Section 2.4.2.1 Appendix B.2 Nuclear Safety Systems and Equipment A list of systems and equipment that ensure the nuclear safety performance criteria can be achieved during and after a plant fire, regardless of fire location, should be developed.

This process can be iterative and can require revisions to incorporate fire risk significant systems and equipment, if further analysis in the circuit analysis or fire area assessment determine additional systems or equipment to be fire risk significant.

The process that follows describes the initial attempt to determine which systems and equipment require evaluation.

Other risk informed-performance-based methods acceptable to the AHJ can be used to refine the list of nuclear safety systems and equipment.

The set of systems and equipment to be considered for nuclear safety should address, as a minimum, the following:.(a) Systems and equipment required to place the plant in a safe and stable condition following a fire occurring while the plant is at power, or while maintaining hot standby or hot shutdown.

This fire also could result in a loss of off-site power, which would require achieving safe and stable conditions using power from on-site ac sources (i.e., emergency diesel generators).

This is typically a traditional Appendix R to 10 CFR 50 post-fire safe shutdown analysis.(b) Systems and equipment required to maintain shutdown cooling capability following a fire originating while the plant is in the shutdown cooling mode.NEI 00-01 Section NEI 00-01 Section Description

3.1.3 Methodology

for Shutdown System Selection 3.1.3.1 Identify Safe Shutdown Functions 3.1.3.2 Identify Combinations of Systems That Satisfy Each Safe Shutdown Function 3.1.3.3 Define Combination of Systems for Each Safe Shutdown Path 3.1.3.4 Assign Shutdown Paths to Each Combination of Systems Refer to Figure 3-2 for a flowchart illustrating the various steps involved in selecting safe shutdown systems and developing the shutdown paths. The following methodology may be used to define the safe shutdown systems and paths for an Appendix R analysis:

[Refer to hard copy of NEI 00-01 for Figure 3-2]Review available documentation to obtain an understanding of the available plant systems and the functions required to achieve and maintain safe shutdown.Documents such as the following may be reviewed:

  • Operating Procedures (Normal, Emergency, Abnormal);
  • System descriptions;
  • Fire Hazard Analysis;
  • Single-line electrical diagrams;
  • Piping and Instrumentation Diagrams (P&IDs)Given the criteria/assumptions defined in Section 3.1.1, identify the available combinations of systems capable of achieving the safe shutdown functions of reactivity control, pressure control, inventory control, decay heat removal, process monitoring and support systems such as electrical and cooling systems (refer to Section 3.1.2). This selection process does not restrict the use of other systems. In addition to achieving the required safe shutdown functions, consider spurious operations and power supply issues that could impact the required safe shutdown function.Select combinations of systems with the capability of performing all of the required safe shutdown functions and designate this set of systems as a safe shutdown path.In many cases, paths may be defined on a divisional basis since the availability of electrical power and other support systems must be demonstrated for each path.During the equipment selection phase, identify any additional support systems and list them for the appropriate path.Assign a path designation to each combination of systems. The path will serve to document the combination of systems relied upon for safe shutdown in each fire area.Refer to Attachment 1 of NEI 00-01 for an example of a table illustrating how to document the various combinations of systems for selected shutdown paths.NFPA 805 Alignment Introductory section, alignment identified in subsections NFPA 805 Reference Aligns Aligns See Nuclear Safety Equipment Technical Report TR08620-015 See Nuclear Safety Equipment Technical Report TR08620-015 Aligns See Nuclear Safety Capability Assessment Technical Report TR08620-312 Aligns See Nuclear Safety Capability Assessment Technical Report TR08620-312 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

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-Methodology Review Page B-9 0 A aeýtdýAttachment B NEPA 805 Section 2.4.2.1 Appendix B.2 Nuclear Safety Systems and Equipment 3.2 Safe Shutdown Equipment Selection The previous section described the methodology for selecting the systems and paths necessary to achieve and maintain safe shutdown for an exposure fire event (see Section 5.0 DEFINITIONS for "Exposure Fire"). This section describes the criteria/assumptions and selection methodology for identifying the specific safe shutdown equipment necessary for the systems to perform their Appendix R function.The selected equipment should be related back to the safe shutdown systems that they support and be assigned to the same safe shutdown path as that system. The list of safe shutdown equipment will then form the basis for identifying the cables necessary for the operation or that can cause the maloperation of the safe shutdown systems.Introductory section, alignment identified in subsections NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-10 Attachment B NFPA 805 Section 2.4.2.1 Appendix B.2.1 Assumptions (Plant Conditions at Time of Postulated Fire)In addition to the assumptions in Chapter 2, the following assumptions apply to this appendix.(a) The plant is in a standard lineup governed by operating procedures, operating modes, or administrative controls at the onset of the fire.(b) Properly oriented check valves function to prevent reverse flow in process systems.(c) Normally closed manual valves (hand-operated only) will remain undamaged by a fire and can be relied upon for system boundary isolation.(d) Instruments located in a fire affected area (e.g., RTDs, thermocouples, pressure transmitters, flow transmitters, and mechanically linked remote/local indications) are assumed to be damaged unless it can be demonstrated otherwise.

The instrument fluid boundary associated with these devices, with the exception of soldered fittings, is assumed to remain intact.(e) Piping, check valves, strainers, tanks, manual valves, heat exchangers, safety relief valves, and pressure vessels are assumed to remain functional during and after a fire.The integrity of instrument tubing, with the exception of soldered fittings, is also expected to be maintained, though the accuracy of the instrument reading can be affected due to heating of the process fluid.NEI 00-01 Section 3.2.1 CriterialAssumptions 3.2.1.1 [A, Primary Components]

3.2.1.1 [B, Secondary Components]

NEI 00-01 Section Description Consider the following criteria and assumptions when identifying equipment necessary to perform the required safe shutdown functions:

Safe shutdown equipment can be divided into two categories.

Equipment may be categorized as (1) primary components or (2) secondary components.

Typically, the following types of equipment are considered to be primary components:

-Pumps, motor operated valves, solenoid valves, fans, gas bottles, dampers, unit coolers, etc. -All necessary process indicators and recorders (i.e., flow indicator, temperature indicator, turbine speed indicator, pressure indicator, level recorder)

-Power supplies or other electrical components that support operation of primary components (i.e., diesel generators, switchgear, motor control centers, load centers, power supplies, distribution panels, etc.).Secondary components are typically items found within the circuitry for a primary component.

These provide a supporting role to the overall circuit function.

Some secondary components may provide an isolation function or a signal to a primary component via either an interlock or input signal processor.

Examples of secondary components include flow switches, pressure switches, temperature switches, level switches, temperature elements, speed elements, transmitters, converters, controllers, transducers, signal conditioners, hand switches, relays, fuses and various instrumentation devices. Determine which equipment should be included on the Safe Shutdown Equipment List (SSEL). As an option, include secondary components with a primary component(s) that would be affected by fire damage to the secondary component.

By doing this, the SSEL can be kept to a manageable size and the equipment included on the SSEL can be readily related to required post-fire safe shutdown systems and functions.

NFPA 805 Alignment Introductory section, alignment identified in subsections Aligns NFPA 805 Reference See Nuclear Safety Equipment Technical Report TR08620-015 Aligns See Nuclear Safety Equipment Technical Report TR08620-015 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

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.~Attachment B NFPA 805 Section 2.4.2.1 Appendix B.2.1 Assumptions (Plant Conditions at Time of Postulated Fire)3.2.1.2 [Fire Damage to Mechanical Components]

3.2.1.3 [Manual Valve Positions]

3.2.1.4 [Check Valves]3.2.1.5 [Instrument Failures]3.2.1.6 [Spurious Components]

3.2.1.7 [Instrument Tubing]Assume that exposure fire damage to manual valves and piping does not adversely impact their ability to perform their pressure boundary or safe shutdown function (heat sensitive piping materials, including tubing with brazed or soldered joints, are not included in this assumption).

Fire damage should be evaluated with respect to the ability to manually open or close the valve should this be necessary as a part of the post-fire safe shutdown scenario.Assume that manual valves are in their normal position as shown on P&IDs or in the plant operating procedures.

Assume that a check valve closes in the direction of potential flow diversion and seats properly with sufficient leak tightness to prevent flow diversion.

Therefore, check valves do not adversely affect the flow rate capability of the safe shutdown systems being used for inventory control, decay heat removal, equipment cooling or other related safe shutdown functions.

Instruments (e.g., resistance temperature detectors, thermocouples, pressure transmitters, and flow transmitters) are assumed to fail upscale, midscale, or downscale as a result of fire damage, whichever is worse. An instrument performing a control function is assumed to provide an undesired signal to the control circuit.Identify equipment that could spuriously operate or mal-operate and impact the performance of equipment on a required safe shutdown path during the equipment selection phase. Consider Bin 1 of RIS 2004-03 during the equipment identification process.Identify instrument tubing that may cause subsequent effects on instrument readings or signals as a result of fire. Determine and consider the fire area location of the instrument tubing when evaluating the effects of fire damage to circuits and equipment in the fire area.Aligns See Nuclear Safety Equipment Technical Report TR08620-015 Aligns Aligns See Nuclear Safety Equipment Technical Report TR08620-015 See Nuclear Safety Equipment Technical Report TR08620-015 Aligns Aligns Aligns See Circuit Analysis Procedure PI 4.4 Appendix F and Technical Report TR07800-009 See Nuclear Safety Equipment Report TR08620-015 and Circuit Analysis TR07800-009.

Technical Report TR08620-019.

See ARC software model for impacts.NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

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..-Attachment B NFPA 805 Section 2.4.2.1 Appendix B.2.2 [Step 2, System Interrelationships]

Step 2: System Interrelationships.

The selection of systems and the documentation of how these systems fulfill the nuclear safety performance criteria should be depicted in system-level logic diagrams, fault trees, or some other method that shows equipment dependencies.

The documentation should consider not only the required process systems but also the essential mechanical/environmental support and essential electrical systems required to support the nuclear safety performance criteria.NEI 00-01 Section 3.1.3.3 Define Combination of Systems for Each Safe Shutdown Path NEI 00-01 Section Description NFPA 805 Alignment NFPA 805 Reference Select combinations of systems with the capability of performing all of the required safe shutdown functions and designate this set of systems as a safe shutdown path.In many cases, paths may be defined on a divisional basis since the availability of electrical power and other support systems must be demonstrated for each path.During the equipment selection phase, identify any additional support systems and list them for the appropriate path.Aligns See Nuclear Safety Capability Assessment Technical Report TR08620-312 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-13 Attachment B NFPA 805 Section 2.4.2.1 Appendix B.2.2 [Step 3(a), Equipment Identification]

Step 3: Equipment Identification.(a) P&IDs (piping and instrumentation diagrams)/flow diagrams should be used to identify the equipment in the flowpath and the boundary equipment within the systems that are required to achieve the nuclear safety objectives.

NEI 00-01 Section 3.2.2.1 Identify the System Flow Path for Each Shutdown Path NEI 00-01 Section Description NFPA 805 Alignment NFPA 805 Reference Mark up and annotate a P&ID to highlight the specific flow paths for each system in support of each shutdown path. Refer to Attachment 2 for an example of an annotated P&ID illustrating this concept. [Refer to hard copy of NEI 00-01 for Attachment A]Aligns See Nuclear Safety Equipment Technical Report TR08620-015 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-14 Attachment B NFPA 805 Section 2.4.2.1 Appendix B.2.2 [Step 3(b), Diversion Equipment](b) Equipment that is not directly in a required system flowpath but whose spurious operation (undesired operation) could prevent achieving the nuclear safety objectives should be identified (e.g., boundary valve component whose spurious opening could divert flow away from critical equipment).

The potential for spurious operations of equipment should be considered when determining boundary valves and equipment selection.

Loops or bypasses within a system where spurious operation would not result in a loss of flow or inadequate flow to nuclear safety success paths need not be considered.

For tanks, all outlet lines should be considered for their functional requirements.

For lines not required to be functional, a means of isolation should be included when necessary to prevent unnecessary drawdown of the tank. Tank fill lines should also be considered.

For example, if two normally closed valves in series must spuriously open to result in an unrecoverable condition, then both valves should be identified on the nuclear safety equipment list (NSEL). If positive means is provided to preclude spurious operation of one valve/component for non-high-low pressure interface component[such as removing power to one of the two motor-operated valves (MOVs) during normal operation], then consideration of the additional component (the other series valve) is not required.NEI 00-01 Section 3.2.2.2 Identify the Equipment in Each Safe Shutdown System Flow Path Including Equipment That May Spuriously Operate and Affect System Operation NEI 00-01 Section Description NFPA 805 Alignment NFPA 805 Reference Review the applicable documentation (e.g. P&IDs, electrical drawings, instrument loop Aligns diagrams) to assure that all equipment in each system's flow path has been identified.

Assure that any equipment that could spuriously operate and adversely affect the desired system function(s) is also identified.

If additional systems are identified which are necessary for the operation of the safe shutdown system under review, include these as systems required for safe shutdown.

Designate these new systems with the same safe shutdown path as the primary safe shutdown system under review (Refer to Figure 3-1). [Refer to hard copy of NEI 00-01 for Figure 3-1]See Nuclear Safety Equipment Technical Report TR08620-015 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

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~G.Attachment B NFPA 805 Section 2.4.2.1 Appendix B.2.2 [Step 3(c), Fire-Induced Plant Transients](c) Careful consideration should be given to equipment that could result in a fire-induced plant transient.

The following is guidance on considerations that should be given in the identification of equipment that could result in a fire-induced plant transient.

(1) Fire-induced plant initiating events [transients and loss of coolant accidents (LOCAs)].Transients are defined as anticipated operational occurrences (e.g., inadvertent safety injection actuation, loss of off-site power, overcooling, overfilling of steam generators, spurious closure of containment isolation valves, significant loss of safety systems, station blackout, rapid cooldown, etc.) that initiate as a result of fire-induced circuit failures.a. Loss of primary system inventory.

The potential for fire initiated spurious actuation at reactor coolant pressure boundaries that could cause an uncontrolled loss of reactor coolant inventory

[e.g., spurious actuation of primary coolant interfaces such as at the reactor head vents, normal and excess letdown at a pressurized water reactor (PWR), main steam relief valves (BWRs)] should be considered.

b. Rapid cooldown.

Transients that could result in an uncontrolled plant cooldown due to spurious operationof boundary valves should be considered.

Interaction of plant systems such as steam generator (PWR) atmospheric dump valves, power-operated relief valves, safety relief valves (BWR) feedwater, reactor trip, turbine trip, and main steam isolation should be considered as well.c. Uncontrolled primary injection.

Transients that could potentially result in an undesired or uncontrolled injection into the reactor coolant system should be assessed.

This can include spurious actuation of high-pressure injection sources (i.e., HPCS, RCIC, HPCI, feedwater for BWRs, high-head ECCS pumps for PWRs).d. Electric power transients.

Transients that could result in a loss of any ac power supplies should be considered.

This loss can include spurious breaker actuations, onsite generating capability spurious starts or failures, or inadvertent paralleling of ac sources due to fire induced circuit failures.NEI 00-01 Section 3.2.1.6 [Spurious Components]

NEI 00-01 Section Description NFPA 805 Alignment NFPA 805 Reference Identify equipment that could spuriously operate or mal-operate and impact the performance of equipment on a required safe shutdown path during the equipment selection phase. Consider Bin 1 of RIS 2004-03 during the equipment identification process.Aligns See Nuclear Safety Equipment Report TR08620-015 and Circuit Analysis TR07800-009.

NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

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.~Attachment B NFPA 805 Section 2.4.2.1 Appendix B.2.2 [Step 3(d), Support Systems](d) Equipment that requires support such as cooling water, instrument air, HVAC, motive power, and control power should be considered in order to understand component and system inter-relationships and sequential equipment loss impact.NEI 00-01 Section 3.2.1.1 [A, Primary Components]

NEI 00-01 Section Description NFPA 805 Alignment NFPA 805 Reference Safe shutdown equipment can be divided into two categories.

Equipment may be categorized as (1) primary components or (2) secondary components.

Typically, the following types of equipment are considered to be primary components:

-Pumps, motor operated valves, solenoid valves, fans, gas bottles, dampers, unit coolers, etc. -All necessary process indicators and recorders (i.e., flow indicator, temperature indicator, turbine speed indicator, pressure indicator, level recorder)

-Power supplies or other electrical components that support operation of primary components (i.e., diesel generators, switchgear, motor control centers, load centers, power supplies, distribution panels, etc.).Aligns See Nuclear Safety Equipment Technical Report TR08620-015 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-17 Attachment B NFPA 805 Section 2.4.2.1 Appendix B.2.2 [Step 3(e), Offsite Power](e) Off-site power can be used as a source of power for nuclear safety equipment.

All equipment required to support the portion of off-site power relied upon to achieve the nuclear safety performance criteria should also be identified.

Off-site power should conservatively be considered available for those cases where availability of off-site power could adversely impact nuclear safety (i.e., reliance cannot be placed on fire causing a loss of off-site power if the consequences of off-site power availability are more severe than its presumed loss). No credit should be taken for a fire causing a loss of offsite power to prevent spurious operations.

NEI 00-01 Section 3.2.1.1 [A, Primary Components]

NEI 00-01 Section Description NFPA 805 Alignment NFPA 805 Reference Safe shutdown equipment can be divided into two categories.

Equipment may be categorized as (1) primary components or (2) secondary components.

Typically, the following types of equipment are considered to be primary components:

-Pumps, motor operated valves, solenoid valves, fans, gas bottles, dampers, unit coolers, etc. -All necessary process indicators and recorders (i.e., flow indicator, temperature indicator, turbine speed indicator, pressure indicator, level recorder)

-Power supplies or other electrical components that support operation of primary components (i.e., diesel generators, switchgear, motor control centers, load centers, power supplies, distribution panels, etc.).Aligns See Nuclear Safety Equipment Technical Report TR08620-015 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

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-Methodology Review Page B-18 A-Attachment B NFPA 805 Section 2.4.2.1 Appendix B.2.2 [Step 3(0), Instrument Sensing Line](f) Instrument sensing lines should be considered for potential inaccurate instrument indications and/or spurious equipment actuations that could occur as a result of an instrument sensing line being exposed to a fire and increased temperatures.

Any instrument sensing lines that could prevent the fulfillment of the nuclear safety performance criteria should be identified, associated with the equipment that it could impact, and included in the nuclear safety assessment for review on a fire area basis.NEI 00-01 Section 3.2.1.7 [Instrument Tubing]NEI 00-01 Section Description NFPA 805 Alignment NFPA 805 Reference Identify instrument tubing that may cause subsequent effects on instrument readings Aligns or signals as a result of fire. Determine and consider the fire area location of the instrument tubing when evaluating the effects of fire damage to circuits and equipment in the fire area.Technical Report TR08620-019.

See ARC software model for impacts.NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

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...m~Attachment B NFPA 805 Section 2.4.2.1 Appendix B.2.2 [Step 3(g), Instrument Air Piping](g) Instrument air piping and components (e.g., accumulators) should be considered for viability during and after the fire in providing the motive force for credited components NEI 00-01 Section NEI 00-01 Section Description NFPA 805 Alignment NFPA 805 Reference 3.2.1.2 [Fire Damage to Assume that exposure fire damage to manual valves and piping does not adversely Mechanical Components]

impact their ability to perform their pressure boundary or safe shutdown function (heat sensitive piping materials, including tubing with brazed or soldered joints, are not included in this assumption).

Fire damage should be evaluated with respect to the ability to manually open or close the valve should this be necessary as a part of the post-fire safe shutdown scenario.Aligns See Nuclear Safety Equipment Technical Report TR08620-015 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

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-Methodology Review Page B-20 Attachment B NFPA 805 Section 2.4.2.1 Appendix B.2.2 [Step 3(h), Power Supplies](h) Power supplies, including alternate power supplies, for nuclear safety equipment should be identified.

Interrelationships between power supplies (such as bus-tie capability and alternate power supplies) should also be identified.

This information is essential in determining nuclear safety equipment losses due to loss of a power supply NEI 00-01 Section 3.2.1.1 [A, Primary Components]

NEI 00-01 Section Description NFPA 805 Alignment NFPA 805 Reference Safe shutdown equipment can be divided into two categories.

Equipment may be categorized as (1) primary components or (2) secondary components.

Typically, the following types of equipment are considered to be primary components:

-Pumps, motor operated valves, solenoid valves, fans, gas bottles, dampers, unit coolers, etc. -All necessary process indicators and recorders (i.e., flow indicator, temperature indicator, turbine speed indicator, pressure indicator, level recorder)

-Power supplies or other electrical components that support operation of primary components (i.e., diesel generators, switchgear, motor control centers, load centers, power supplies, distribution panels, etc.).Aligns See Nuclear Safety Equipment Technical Report TR08620-015 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

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-Methodology Review Page B-21 C' 4 Attachment B NFPA 805 Section 2.4.2.1 Appendix B.2.2 [Step 4, Equipment Interrelationships]

Step 4: Equipment Interrelationships.

The necessary relationships between individual nuclear safety equipment and systems should be understood and documented.

NEI 00-01 Section NEI 00-01 Section Description NFPA 805 Alignment NFPA 805 Reference 3.2.1.5 [Instrument Instruments (e.g., resistance temperature detectors, thermocouples, pressure Aligns See Circuit Analysis Procedure PI 4.4 Appendix F Failures) transmitters, and flow transmitters) are assumed to fail upscale, midscale, or and Technical Report TR07800-009 downscale as a result of fire damage, whichever is worse. An instrument performing a control function is assumed to provide an undesired signal to the control circuit.NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

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-Methodology Review Page B-22 Attachment B NFPA 805 Section 2.4.2.1 Appendix B.2.2 [Step 5(a), Equipment Selection Documentation]

Step 5: Documentation.(a) The bases for selection and exclusion of nuclear safety systems and equipment should be documented and maintained.

Calculations and analyses that have been previously performed in support of other nuclear safety objectives (i.e., station blackout, seismic qualification) can be utilized provided the results of these analyses have properly considered the applicability to post-fire nuclear safety.NEI 00-01 Section 3.2.2.3 Develop a List of Safe Shutdown Equipment and Assign the Corresponding System and Safe Shutdown Path(s)Designation to Each.3.2.2.4 Identify Equipment Information Required for the Safe Shutdown Analysis 3.2.2.5 Identify Dependencies Between Equipment, Supporting Equipment, Safe Shutdown Systems and Safe Shutdown Paths.NEI 00-01 Section Description NFPA 805 Alignment NFPA 805 Reference Prepare a table listing the equipment identified for each system and the shutdown path that it supports.

Identify any valves or other equipment that could spuriously operate and impact the operation of that safe shutdown system. Assign the safe shutdown path for the affected system to this equipment.

During the cable selection phase, identify additional equipment required to support the safe shutdown function of the path (e.g., electrical distribution system equipment).

Include this additional equipment in the safe shutdown equipment list. Attachment 3 to this document provides an example of a (SSEL). The SSEL identifies the list of equipment within the plant considered for safe shutdown and it documents various equipment-related attributes used in the analysis.[Refer to hard copy of NEI 00-01 for Attachment 3]Collect additional equipment-related information necessary for performing the post-fire safe shutdown analysis for the equipment.

In order to facilitate the analysis, tabulate this data for each piece of equipment on the SSEL. Refer to Attachment 3 to this document for an example of a SSEL. Examples of related equipment data should include the equipment type, equipment description, safe shutdown system, safe shutdown path, drawing reference, fire area, fire zone, and room location of equipment.

Other information such as the following may be useful in performing the safe shutdown analysis:

normal position, hot shutdown position, cold shutdown position, failed air position, failed electrical position, high/low pressure interface concem, and spurious operation concem. [Refer to hard copy of NEI 00-01 for Attachment 3]Aligns Aligns See Nuclear Safety Equipment Technical Report TR08620-015 See Nuclear Safety Equipment Technical Report TR08620-015 See Circuit analysis Project Instruction PI 4.4 and Technical Report TR07800-009 In the process of defining equipment and cables for safe shutdown, identify additional Aligns supporting equipment such as electrical power and interlocked equipment.

As an aid in assessing identified impacts to safe shutdown, consider modeling the dependency between equipment within each safe shutdown path either in a relational database or in the form of a Safe Shutdown Logic Diagram (SSLD). Attachment 4 provides an example of a SSLD that may be developed to document these relationships.

[Refer to hard copy of NEI 00-01 for Attachment 4]NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-23 A oý Attachment B NFPA 805 Section 2.4.2.1 Appendix B.2.2 Considerations for the Selection of Nuclear Safety Systems and Equipment.

[Step 1]Step 1: System Identification.

Based upon documentation of plant design, risk insights, and operation, plant systems required to achieve each of the nuclear safety criteria should be identified.

NEI 00-01 Section NEI 00-01 Section Description NFPA 805 Alignment NFPA 805 Reference 3.2.2 Methodology for Refer to Figure 3-3 for a flowchart illustrating the various steps involved in selecting Aligns See Nuclear Safety Equipment Technical Report Equipment Selection safe shutdown equipment.

Use the following methodology to select the safe shutdown TR08620-015 equipment for a post-fire safe shutdown analysis:

[Refer to hard copy of NEI 00-01 for Figure 3-3]NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-24 Attachment B NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis 2.4.2.2.1 Circuits Required in Nuclear Safety Functions.

Circuits required for the nuclear safety functions shall be identified.

This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. 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. This will ensure that a comprehensive population of circuitry is evaluated.(See Appendix B for considerations in analyzing circuits.)

2.4.2.2.2 Other Required Circuits.Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.(a) Common Power Supply Circuits.

Those circuits whose fire induced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified.

This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.(See Appendix B for considerations when analyzing common power supply concerns.)

NEI 00-01 Section NEI 00-01 Section Description 3.3 Safe Shutdown Cable Selection and Location 3.3.1 Criteria/Assumptions This section provides industry guidance on the recommended methodology and criteria for selecting safe shutdown cables and determining their potential impact on equipment required for achieving and maintaining safe shutdown of an operating nuclear power plant for the condition of an exposure fire. The Appendix R safe shutdown cable selection criteria are developed to ensure that all cables that could affect the proper operation or that could cause the maloperation of safe shutdown equipment are identified and that these cables are properly related to the safe shutdown equipment whose functionality they could affect. Through this cable-to-equipment relationship, cables become part of the safe shutdown path assigned to the equipment affected by the cable.To identify an impact to safe shutdown equipment based on cable routing, the equipment must have cables that affect it identified.

Carefully consider how cables are related to safe shutdown equipment so that impacts from these cables can be properly assessed in terms of their ultimate impact on safe shutdown system equipment.

Consider the following criteria when selecting cables that impact safe shutdown equipment:

NFPA 805 Alignment Introductory section, alignment identified in subsections Introductory section, alignment identified in subsections NFPA 805 Reference NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-25

._C Attachment B NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis 3.3.1.1 [Cable Selection]

3.3.1.2 [Cables Affecting Multiple Components]

3.3.1.3 [Isolation devices]3.3.1.4 [Identify "Not Required" Cables]3.3.1.5 [Identify Power Supplies]3.3.1.6 [ESFAS Actuation]

The list of cables whose failure could impact the operation of a piece of safe shutdown equipment includes more than those cables connected to the equipment.

The relationship between cable and affected equipment is based on a review of the electrical or elementary wiring diagrams.

To assure that all cables that could affect the operation of the safe shutdown equipment are identified, investigate the power, control, instrumentation, interlock, and equipment status indication cables related to the equipment.

Consider reviewing additional schematic diagrams to identify additional cables for interlocked circuits that also need to be considered for their impact on the ability of the equipment to operate as required in support of post fire safe shutdown.As an option, consider applying the screening criteria from Section 3.5 as a part of this section. For an example of this see Section 3.3.1.4.In cases where the failure (including spurious actuations) of a single cable could impact more than one piece of safe shutdown equipment, include the cable with each piece of safe shutdown equipment.

Electrical devices such as relays, switches and signal resistor units are considered to be acceptable isolation devices. In the case of instrument loops, review the isolation capabilities of the devices in the loop to determine that an acceptable isolation device has been installed at each point where the loop must be isolated so that a fault would not impact the performance of the safe shutdown instrument function.Screen out cables for circuits that do not impact the safe shutdown function of a component (i.e., annunciator circuits, space heater circuits and computer input circuits) unless some reliance on these circuits is necessary.

However, they must be isolated from the component's control scheme in such a way that a cable fault would not impact the performance of the circuit.For each circuit requiring power to perform its safe shutdown function, identify the cable supplying power to each safe shutdown and/or required interlock component.

Initially, identify only the power cables from the immediate upstream power source for these interlocked circuits and components (i.e., the closest power supply, load center or motor control center). Review further the electrical distribution system to capture the remaining equipment from the electrical power distribution system necessary to support delivery of power from either the offsite power source or the emergency diesel generators (i.e., onsite power source) to the safe shutdown equipment.

Add this equipment to the safe shutdown equipment list. Evaluate the power cables for this additional equipment for associated circuits concerns.The automatic initiation logics for the credited post-fire safe shutdown systems are not required to support safe shutdown.

Each system can be controlled manually by operator actuation in the main control room or emergency control station. If operator actions outside the MCR are necessary, those actions must conform to the regulatory requirements on manual actions. However, if not protected from the effects of fire, the fire-induced failure of automatic initiation logic circuits must not adversely affect any post-fire safe shutdown system function.Aligns See Circuit analysis Project Instruction PI 4.4 and Technical Report TR07800-009 Aligns Aligns See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 See Circuit analysis Project Instruction PI 4.4 and Technical Report TR07800-009 See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 Aligns Aligns Aligns See MSO Technical Report TR08620-025, and Nuclear Safety Capability Assessment Technical Report TR08620-312 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-26 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-26 Attachment B NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis 3.3.1.7 [Circuit Coordination]

3.3.2 Associated

Circuit Cables 3.3.2 [A, Cables Whose Failure May Cause Spurious Actuations]

3.3.2 [B, Common Power Source Cables]3.3.2. [C, Common Enclosure Cables]Cabling for the electrical distribution system is a concern for those breakers that feed associated circuits and are not fully coordinated with upstream breakers.

With respect to electrical distribution cabling, two types of cable associations exist. For safe shutdown considerations, the direct power feed to a primary safe shutdown component is associated with the primary component.

For example, the power feed to a pump is necessary to support the pump. Similarly, the power feed from the load center to an MCC supports the MCC. However, for cases where sufficient branch-circuit coordination is not provided, the same cables discussed above would also support the power supply. For example, the power feed to the pump discussed above would support the bus from which it is fed because, for the case of a common power source analysis, the concern is the loss of the upstream power source and not the connected load. Similarly, the cable feeding the MCC from the load center would also be necessary to support the load center.Appendix R,Section III.G.2, requires that separation features be provided for equipment and cables, including associated nonsafety circuits that could prevent operation or cause maloperation due to hot shorts, open circuits, or shorts to ground, of redundant trains of systems necessary to achieve hot shutdown.

The three types of associated circuits were identified in Reference 6.1.5 and further clarified in a NRC memorandum dated March 22, 1982 from R. Mattson to D. Eisenhut, Reference 6.1.6. They are as follows: Spurious actuations; Common power source; Common enclosure.

Safe shutdown system spurious actuation concerns can result from fire damage to a cable whose failure could cause the spurious actuation/mal-operation of equipment whose operation could affect safe shutdown.

These cables are identified in Section 3.3.3 together with the remaining safe shutdown cables required to support control and operation of the equipment.

The concern for the common power source associated circuits is the loss of a safe shutdown power source due to inadequate breaker/fuse coordination.

In the case of a fire-induced cable failure on a non-safe shutdown load circuit supplied from the safe shutdown power source, a lack of coordination between the upstream supply breaker/fuse feeding the safe shutdown power source and the load breaker/fuse supplying the non-safe shutdown faulted circuit can result in loss of the safe shutdown bus. This would result in the loss of power to the safe shutdown equipment supplied from that power source preventing the safe shutdown equipment from performing its required safe shutdown function.

Identify these cables together with the remaining safe shutdown cables required to support control and operation of the equipment.

Refer to Section 3.5.2.4 for an acceptable methodology for analyzing the impact of these cables on post-fire safe shutdown.The concern with common enclosure associated circuits is fire damage to a cable whose failure could propagate to other safe shutdown cables in the same enclosure either because the circuit is not properly protected by an isolation device (breaker/fuse) such that a fire-induced fault could result in ignition along its length, or by the fire propagating along the cable and into an adjacent fire area. This fire spread to an adjacent fire area could impact safe shutdown equipment in that fire area, thereby resulting in a condition that exceeds the criteria and assumptions of this methodology (i.e., multiple fires). Refer to Section 3.5.2.5 for an acceptable methodology for analyzing the impact of these cables on post-fire safe shutdown.Aligns See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 Aligns See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 Aligns See Nuclear Safety Capability Assessment Technical Report TR08620-312 See Circuit analysis Project Instruction PI 4.4 and Technical Report TR07800-009 Aligns Aligns See Circuit analysis Project Instruction PI 4.4 and Technical Report TR07800-009 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-27

~G.Attachment B NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis 3.3.3 Methodology for Cable Selection and Location 3.3.3.1 Identify Circuits Required for the Operation of the Safe Shutdown Equipment 3.3.3.2 Identify Interlocked Circuits and Cables Whose Spurious Operation or Mal-operation Could Affect Shutdown 3.3.3.3 Assign Cables to the Safe Shutdown Equipment 3.3.3.4 Identify Routing of Cables 3.3,3.5 Identify Location of Raceway and Cables by Fire Area Refer to Figure 3-4 for a flowchart illustrating the various steps involved in selecting the cables necessary for performing a post-fire safe shutdown analysis.

Use the following methodology to define the cables required for safe shutdown including cables that may cause associated circuits concerns for a post-fire safe shutdown analysis:

[Refer to hard copy of NEI 00-01 for Figure 3-4]For each piece of safe shutdown equipment defined in section 3.2, review the appropriate electrical diagrams including the following documentation to identify the circuits (power, control, instrumentation) required for operation or whose failure may impact the operation of each piece of equipment:

Single-line electrical diagrams;Elementary wiring diagrams;

-Electrical connection diagrams Instrument loop diagrams.

For electrical power distribution equipment such as power supplies, identify any circuits whose failure may cause a coordination concern for the bus under evaluation.

If power is required for the equipment, include the closest upstream power distribution source on the safe shutdown equipment list. Through the iterative process described in Figures 3-2 and 3-3, include the additional upstream power sources up to either the offsite or the emergency power source. [Refer to hard copy of NEI 00-01 for Figure 3-2 and 3-3]In reviewing each control circuit, investigate interlocks that may lead to additional circuit schemes, cables and equipment.

Assign to the equipment any cables for interlocked circuits that can affect the equipment.

While investigating the interlocked circuits, additional equipment or power sources may be discovered.

Include these interlocked equipment or power sources in the safe shutdown equipment list (refer to Figure 3-3) if they can impact the operation of the equipment under consideration.

[Refer to hard copy of NEI 00-01 for Figure 3-3]Given the criteria/assumptions defined in Section 3.3.1, identify the cables required to operate or that may result in maloperation of each piece of safe shutdown equipment.

Tabulate the list of cables potentially affecting each piece of equipment in a relational database including the respective drawing numbers, their revision and any interlocks that are investigated to determine their impact on the operation of the equipment.

In certain cases, the same cable may support multiple pieces of equipment.

Relate the cables to each piece of equipment, but not necessarily to each supporting secondary component.

If adequate coordination does not exist for a particular circuit, relate the power cable to the power source. This will ensure that the power source is identified as affected equipment in the fire areas where the cable may be damaged.Identify the routing for each cable including all raceway and cable endpoints.

Typically, this information is obtained from joining the list of safe shutdown cables with an existing cable and raceway database.Identify the fire area location of each raceway and cable endpoint identified in the previous step and join this information with the cable routing data. In addition, identify the location of field-routed cable by fire area. This produces a database containing all of the cables requiring fire area analysis, their locations by fire area, and their raceway.Introductory section, alignment identified in subsections Aligns See Circuit analysis Project Instruction PI 4.4 and Technical Report TR07800-009.

Results of required circuits/routing documented in PC-CKS Database.Aligns See Circuit analysis Project Instruction PI 4.4 and Technical Report TR07800-009 Aligns See Circuit analysis Project Instruction PI 4.4 and Technical Report TR07800-009 Aligns Aligns See Circuit analysis Project Instruction PI 4.4 and Technical Report TR07800-009 See Circuit analysis Project Instruction PI 4.4 and Technical Report TR07800-009.

Circuit Routing maintained in PC-CKS Database.NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-28 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-28 AýtýAttachment B NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis 3.5 Circuit Analysis and Evaluation

3.5.1 CriterialAssumptions

3.5.1.1 [Circuit Failure Types and Its Impact]3.5.1.2 [Circuit Contacts and Operational Modes]3.5.1.3 [Duration of Circuit Failures]3.5.1.4 [Cable Failure Configurations]

This section on circuit analysis provides information on the potential impact of fire on circuits used to monitor, control and power safe shutdown equipment.

Applying the circuit analysis criteria will lead to an understanding of how fire damage to the cables may affect the ability to achieve and maintain post-fire safe shutdown in a particular fire area. This section should be used in conjunction with Section 3.4, to evaluate the potential fire-induced impacts that require mitigation.

Appendix R Section III.G.2 identifies the fire-induced circuit failure types that are to be evaluated for impact from exposure fires on safe shutdown equipment.

Section III.G.2 of Appendix R requires consideration of hot shorts, shorts-to-ground and open circuits.Apply the following criteria/assumptions when performing fire-induced circuit failure evaluations.

Consider the following circuit failure types on each conductor of each unprotected safe shutdown cable to determine the potential impact of a fire on the safe shutdown equipment associated with that conductor.

A hot short may result from a fire-induced insulation breakdown between conductors of the same cable, a different cable or from some other external source resulting in a compatible but undesired impressed voltage or signal on a specific conductor.

A hot short may cause a spurious operation of safe shutdown equipment.

An open circuit may result from a fire-induced break in a conductor resulting in the loss of circuit continuity.

An open circuit may prevent the ability to control or power the affected equipment.

An open circuit may also result in a change of state for normally energized equipment. (e.g. [for BWRs] loss of power to the Main Steam Isolation Valve (MSIV) solenoid valves due to an open circuit will result in the closure of the MSIVs). Note that RIS 2004-03 indicates that open circuits, as an initial mode of cable failures, are considered to be of very low likelihood.

The risk-informed inspection process will focus on failures with relatively high probabilities.

A short-to-ground may result from a fire-induced breakdown of a cable insulation system, resulting in the potential on the conductor being applied to ground potential.

A short-to-ground may have all of the same effects as an open circuit and, in addition, a short-to-ground may also cause an impact to the control circuit or power train of which it is a part. Consider the three types of circuit failures identified above to occur individually on each conductor of each safe shutdown cable on the required safe shutdown path in the fire area.Assume that circuit contacts are positioned (i.e., open or closed) consistent with the normal mode/position of the safe shutdown equipment as shown on the schematic drawings.

The analyst must consider the position of the safe shutdown equipment for each specific shutdown scenario when determining the impact that fire damage to a particular circuit may have on the operation of the safe shutdown equipment.

Assume that circuit failure types resulting in spurious operations exist until action has been taken to isolate the given circuit from the fire area, or other actions have been taken to negate the effects of circuit failure that is causing the spurious actuation.

The fire is not assumed to eventually clear the circuit fault. Note that RIS 2004-03 indicates that fire-induced hot shorts typically self-mitigate after a limited period of time.When both trains are in the same fire area outside of primary containment, all cables that do not meet the separation requirements of Section III.G.2 are assumed to fail in their worst case configuration.

Introductory section, alignment identified in subsections Introductory section, alignment identified in subsections Aligns See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 Aligns Aligns See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 Aligns NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-29 SCS&G*A geý09ýAttachment B NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis 3.5.1.5 [A, Circuit Failure General Guidance]3.5.1.5 [A, Circuit Failure Risk Assessment Guide]3.5.1.5 [B, Cable Failure Modes]3.5.1.5 [C, Multiple Cable Damage]3.5.1.5 [D, DC circuits]The following guidance provides the NRC inspection focus from Bin 1 of RIS 2004-03 in order to identify any potential combinations of spurious operations with higher risk significance.

Bin 1 failures should also be the focus of the analysis; however, NRC has indicated that other types of failures required by the regulations for analysis should not be disregarded even if in Bin 2 or 3. If Bin 1 changes in subsequent revisions of RIS 2004-03, the guidelines in the revised RIS should be followed.Cable Failure Modes. For multiconductor cables testing has demonstrated that conductor-to-conductor shorting within the same cable is the most common mode of failure. This is often referred to as "intra-cable shorting." It is reasonable to assume that given damage, more than one conductor-to-conductor short will occur in a given cable. A second primary mode of cable failure is conductor-to-conductor shorting between separate cables, commonly referred to as "inter-cable shorting." Inter-cable shorting is less likely than intra-cable shorting.

Consistent with the current knowledge of fire-induced cable failures, the following configurations should be considered:

For any individual multiconductor cable (thermoset or thermoplastic), any and all potential spurious actuations that may result from intra-cable shorting, including any possible combinationof conductors within the cable, may be postulated to occur concurrently regardless of number. However, as a practical matter, the number of combinations of potential hot shorts increases rapidly with the number of conductors within a given cable. For example, a multiconductor cable with three conductors (3C) has 3 possible combinations of two (including desired combinations), while a five conductor cable (5C) has 10 possible combinations of two (including desired combinations), and a seven conductor cable (7C) has 21 possible combinations of two (including desired combinations).

To facilitate an inspection that considers most of the risk presented by postulated hot shorts within a multiconductor cable, inspectors should consider only a few (three or four) of the most critical postulated combinations.

For any thermoplastic cable, any and all potential spurious actuations that may result from intra-cable and inter-cable shorting with other thermoplastic cables, including any possible combination of conductors within or between the cables, may be postulated to occur concurrently regardless of number. (The consideration of thermoset cable inter-cable shorts is deferred pending additional research.)

For cases involving the potential damage of more than one multiconductor cable, a maximum of two cables should be assumed to be damaged concurrently.

The spurious actuations should be evaluated as previously described.

The consideration of more than two cables being damaged (and subsequent spurious actuations) is deferred pending additional research.For cases involving direct current (DC) circuits, the potential spurious operation due to failures of the associated control cables (even if the spurious operation requires two concurrent hot shorts of the proper polarity, e.g., plus-to-plus and minus-to-minus) should be considered when the required source and target conductors are each located within the same multiconductor cable.Aligns Aligns See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 Aligns Aligns See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 Aligns NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-30 "lewsm Attachment B NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis 3.5.1.5 [E, Instrumentation Circuits]3.5.1.5 [F, Undesired Consequences]

3.5.2 Types

of Circuit Failures Instrumentation Circuits.

Required instrumentation circuits are beyond the scope of this associated circuit approach and must meet the same requirements as required power and control circuits.

There is one case where an instrument circuit could potentially be considered an associated circuit. If fire-induced damage of an instrument circuit could prevent operation (e.g., lockout permissive signal) or cause maloperation (e.g., unwanted start/stop/reposition signal) of systems necessary to achieve and maintain hot shutdown, then the instrument circuit may be considered an associated circuit and handled accordingly.

Likelihood of Undesired Consequences.

Determination of the potential consequence of the damaged associated circuits is based on the examination of specific NPP piping and instrumentation diagrams (P&IDs) and review of components that could prevent operation or cause maloperation such as flow diversions, loss of coolant, or other scenarios that could significantly impair the NPP's ability to achieve and maintain hot shutdown.

When considering the potential consequence of such failures, the [analyst]should also consider the time at which the prevented operation or maloperation occurs. Failures that impede hot shutdown within the first hour of the fire tend to be most risk significant in a first-order evaluation.

Consideration of cold-shutdown circuits is deferred pending additional research.Appendix R requires that nuclear power plants must be designed to prevent exposure fires from defeating the ability to achieve and maintain post-fire safe shutdown.

Fire damage to circuits that provide control and power to equipment on the required safe shutdown path and any other equipment whose spurious operation/mal-operation could affect shutdown in each fire area must be evaluated for the effects of a fire in that fire area. Only one fire at a time is assumed to occur. The extent of fire damage is assumed to be limited by the boundaries of the fire area. Given this set of conditions, it must be assured that one redundant train of equipment capable of achieving hot shutdown is free of fire damage for fires in every plant location.

To provide this assurance, Appendix R requires that equipment and circuits required for safe shutdown be free of fire damage and that these circuits be designed for the fire-induced effects of a hot short, short-to-ground, and open circuit. With respect to the electrical distribution system, the issue of breaker coordination must also be addressed.

This section will discuss specific examples of each of the following types of circuit failures:

Open circuit; Short-to-ground; Hot short.Aligns See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 Aligns See NSCA ARC software model, results documented in Nuclear Safety Capability Assessment Technical Report TR08620-312 Introductory section, alignment identified in subsections NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-31 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-31 Attachment B NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis 3.5.2.1 Circuit Failures Due to an Open Circuit 3.5.2.2 Circuit Failures Due to a Short-to-Ground

[A, General]3.5.2.2 Circuit Failures Due to a Short-to-Ground

[B, Grounded Circuits]This section provides guidance for addressing the effects of an open circuit for safe shutdown equipment.

An open circuit is a fire-induced break in a conductor resulting in the loss of circuit continuity.

An open circuit will typically prevent the ability to control or power the affected equipment.

An open circuit can also result in a change of state for normally energized equipment.

For example, a loss of power to the main steam isolation valve (MSIV) solenoid valves [for BWRs] due to an open circuit will result in the closure of the MSIV. NOTE: The EPRI circuit failure testing indicated that open circuits are not likely to be the initial fire-induced circuit failure mode. Consideration of this may be helpful within the safe shutdown analysis.

Consider the following consequences in the safe shutdown circuit analysis when determining the effects of open circuits:

Loss of electrical continuity may occur within a conductor resulting in deenergizing the circuit and causing a loss of power to, or control of, the required safe shutdown equipment.

In selected cases, a loss of electrical continuity may result in loss of power to an interlocked relay or other device. This loss of power may change the state of the equipment.

Evaluate this to determine if equipment fails safe. Open circuit on a high voltage (e.g., 4.16 kV) ammeter current transformer (CT) circuit may result in secondary damage. [Refer to hard copy of NEI 00-01 for Figure 3.5.2-1]Open circuit No. 1: An open circuit at location No. 1 will prevent operation of the subject equipment.

Open circuit No. 2: An open circuit at location No. 2 will prevent opening/starting of the subject equipment, but will not impact the ability to close/stop the equipment.

This section provides guidance for addressing the effects of a short-to-ground on circuits for safe shutdown equipment.

A short-to-ground is a fire-induced breakdown of a cable insulation system resulting in the potential on the conductor being applied to ground potential.

A short-to-ground can cause a loss of power to or control of required safe shutdown equipment.

In addition, a short-to-ground may affect other equipment in the electrical power distribution system in the cases where proper coordination does not exist. Consider the following consequences in the post-fire safe shutdown analysis when determining the effects of circuit failures related to shorts-to-ground:

-A short to ground in a power or a control circuit may result in tripping one or more isolation devices (i.e. breaker/fuse) and causing a loss of power to or control of required safe shutdown equipment.

-In the case of certain energized equipment such as HVAC dampers, a loss of control power may result in loss of power to an interlocked relay or other device that may cause one or more spurious operations.

Typically, in the case of a grounded circuit, a short-to-ground on any part of the circuit would present a concern for tripping the circuit isolation device thereby causing a loss of control power. Figure 3.5.2-2 illustrates how a short-to-ground fault may impact a grounded circuit. Short-to-ground No. 1: A short-to-ground at location No. 1 will result in the control power fuse blowing and a loss of power to the control circuit. This will result an inability to operate the equipment using the control switch. Depending on the coordination characteristics between the protective device on this circuit and upstream circuits, the power supply to other circuits could be affected.

Short-to-ground No. 2: A short-to-ground at location No. 2 will have no effect on the circuit until the close/stop control switch is closed. Should this occur, the effect would be identical to that for the short-to-ground at location No. 1 described above. Should the open/start control switch be closed prior to closing the close/stop control switch, the equipment will still be able to be opened/started.

[Refer to hard copy of NEI 00-01 for Figure 3.5.2-2]Aligns Aligns See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 Aligns See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-32

  • ~Attachment B NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis 3.5.2.2 Circuit Failures Due to a Short-to-Ground

[C, Ungrounded Circuits]3.5.2.3 Circuit Failures Due to a Hot Short [A, General]3.5.2.3 Circuit Failures Due to a Hot Short [B, Grounded Circuits]In the case of an ungrounded circuit, postulating only a single short-to-ground on any part of the circuit may not result in tripping the circuit isolation device. Another short-to-ground on the circuit or another circuit from the same source would need to exist to cause a loss of control power to the circuit. Figure 3.5.2-3 illustrates how a short to ground fault may impact an ungrounded circuit. Short-to-ground No. 1: A short-to-ground at location No. I will result in the control power fuse blowing and a loss of power to the control circuit if short-to-ground No. 3 also exists either within the same circuit or on any other circuit fed from the same power source. This will result in an inability to operate the equipment using the control switch. Depending on the coordination characteristics between the protective device on this circuit and upstream circuits, the power supply to other circuits could be affected.

Short-to-ground No. 2: A short-to-ground at location No. 2 will have no effect on the circuit until the close/stop control switch is closed. Should this occur, the effect would be identical to that for the short-to-ground at location No. 1 described above. Should the open/start control switch be closed prior to closing the close/stop control switch, the equipment will still be able to be opened/started.

[Refer to hard copy of NEI 00-01 for Figure 3.5.2-3]This section provides guidance for analyzing the effects of a hot short oncircuits for required safe shutdown equipment.

A hot short is defined as a fire induced insulation breakdown between conductors of the same cable, a different cable or some other external source resulting in an undesired impressed voltage on a specific conductor.

The potential effect of the undesired impressed voltage would be to cause equipment to operate or fail to operate in an undesired manner. Consider the following specific circuit failures related to hot shorts as part of the post-fire safe shutdown analysis:

-A hot short between an energized conductor and a de-energized conductor within the same cable may cause a spurious actuation of equipment.

The spuriously actuated device (e.g., relay) may be interlocked with another circuit that causes the spurious actuation of other equipment.

This type of hot short is called a conductor-to-conductor hot short or an internal hot short. -A hot short between any external energized source such as an energized conductor from another cable (thermoplastic cables only)and a de-energized conductor may also cause a spurious actuation of equipment.

This is called a cable-to-cable hot short or an external hot short. Cable-to-cable hot shorts between thermoset cables are not postulated to occur pending additionalresearch.

A short-to-ground is another failure mode for a grounded control circuit. A short-to-ground as described above would result in de-energizing the circuit. This would further reduce the likelihood for the circuit to change the state of the equipment either from a control switch or due to a hot short. Nevertheless, a hot short still needs to be considered.

Figure 3.5.2-4 shows a typical grounded control circuit that might be used for a motor-operated valve. However, the protective devices and position indication lights that would normally be included in the control circuit for a motor-operated valve have been omitted, since these devices are not required to understand the concepts being explained in this section. In the discussion provided below, it is assumed that a single fire in a given fire area could cause any one of the hot shorts depicted.

The following discussion describes how to address the impact of these individual cable faults on the operation of the equipment controlled by this circuit. Hot short No. 1: A hot short at this location would energize the close relay and result in the undesired closure of a motor-operated valve. Hot short No. 2: A hot short at this location would energize the open relay and result in the undesired opening of a motor-operated valve. [Refer to hard copy of NEI 00-01 for Figure 3.5.2-4]Aligns See NFPA and Fire PRA Circuit Analysis Technical Report TR07800-009 Aligns See Circuit analysis Project Instruction PI 4.4 and Technical Report TR07800-009 Aligns See Circuit analysis Project Instruction PI 4.4 and Technical Report TR07800-009 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-33 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-33 Attachment B NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis 3.5.2.3 Circuit Failures Due to a Hot Short [C, Ungrounded Circuits]In the case of an ungrounded circuit, a single hot short may be sufficient to cause a spurious operation.

A single hot short can cause a spurious operation if the hot short comes from a circuit from the positive leg of the same ungrounded source as the affected circuit. In reviewing each of these cases, the common denominator is that in every case, the conductor in the circuit between the control switch and the start/stop coil must be involved.

Figure 3.5.2-5 depicted below {see Figure in NEI 00-01, Rev 1]shows a typical ungrounded control circuit that might be used for a motor-operated valve. However, the protective devices and position indication lights that would normally be included in the control circuit for a motor-operated valve have been omitted, since these devices are not required to understand the concepts being explained in this section. In the discussion provided below, it is assumed that a single fire in a given fire area could cause any one of the hot shorts depicted.

The discussion provided below describes how to address the impact of these cable faults on the operation of the equipment controlled by this circuit. Hot short No. 1: A hot short at this location from the same control power source would energize the close relay and result in the undesired closure of a motor operated valve. Hot short No. 2: A hot short at this location from the same control power source would energize the open relay and result in the undesired opening of a motor operated valve. [Refer to hard copy of NEI 00-01 for Figure 3.5.2-5]Aligns See Circuit analysis Project Instruction PI 4.4 and Technical Report TR07800-009 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-34 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-34 SCOAGO.AdtdýAttachment B NFPA 805 Section 2.4.2.3* Nuclear Safety Equipment and Cable Location.Physical location of equipment and cables shall be identified. (See Appendix B for considerations when identifying locations.)

[Note: A.2.4.2.3 Equipment and cables should be located by the smallest designator (room, fire zone, or fire area) for ease of analysis.]

NEI 00-01 Section NEI 00-01 Section Description NFPA 805 Alignment NFPA 805 Reference 3.5.2.4 Circuit Failures Due to Inadequate Circuit Coordination

[A, General]3.5.2.4 Circuit Failures Due to Inadequate Circuit Coordination

[A, Methodology]

The evaluation of associated circuits of a common power source consists of verifying proper coordination between the supply breaker/fuse and the load breakers/fuses for power sources that are required for safe shutdown.

The concern is that, for fire damage to a single power cable, lack of coordination between the supply breaker/fuse and the load breakers/fuses can result in the loss of power to a safe shutdown power source that is required to provide power to safe shutdown equipment.

For the example shown in Figure 3.5.2-6, the circuit powered from load breaker 4 supplies power to a non-safe shutdown pump. This circuit is damaged by fire in the same fire area as the circuit providing power to from the Train B bus to the Train B pump, which is redundant to the Train A pump. To assure safe shutdown for a fire in this fire area, the damage to the non-safe shutdown pump powered from load breaker 4 of the Train A bus cannot impact the availability of the Train A pump, which is redundant to the Train B pump. To assure that there is no impact to this Train A pump due to the associated circuits' common power source breaker coordination issue, load breaker 4 must be fully coordinated with the feeder breaker to the Train A bus. A coordination study should demonstrate the coordination status for each required common power source. For coordination to exist, the time-current curves for the breakers, fuses and/or protective relaying must demonstrate that a fault on the load circuits is isolated before tripping the upstream breaker that supplies the bus. Furthermore, the available short circuit current on the load circuit must be considered to ensure that coordination is demonstrated at the maximum fault level. [Refer to hard copy of NEI 00-01 for Figure 3.5.2-6]The methodology for identifying potential associated circuits of a common power source and evaluating circuit coordination cases of associated circuits on a single circuit fault basis is as follows: -Identify the power sources required to supply power to safe shutdown equipment.

-For each power source, identify the breaker/fuse ratings, types, trip settings and coordination characteristics for the incoming source breaker supplying the bus and the breakers/fuses feeding the loads supplied by the bus. -For each power source, demonstrate proper circuit coordination using acceptable industry methods. -For power sources not properly coordinated, tabulate by fire area the routing of cables whose breaker/fuse is not properly coordinated with the supply breaker/fuse.

Evaluate the potential for disabling power to the bus in each of the fire areas in which the associated circuit cables of concern are routed and the power source is required for safe shutdown.

Prepare a list of thefollowing information for each fire area: -Cables of concem. -Affected common power source and its path. Raceway in which the cable is enclosed.

-Sequence of the raceway in the cable route.- Fire zone/area in which the raceway is located. For fire zones/areas in which the power source is disabled, the effects are mitigated by appropriate methods.Develop analyzed safe shutdown circuit dispositions for the associated circuit of concern cables routed in an area of the same path as required by the power source.Evaluate adequate separation based upon the criteria in Appendix R, NRC staff guidance, and plant licensing bases.Aligns See Circuit analysis Project Instruction PI 4.4 and Technical Report TR07800-009 Aligns See Circuit analysis Project Instruction PI 4.4 and Technical Report TR07800-009 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-35 Attachment B NFPA 805 Section 2.4.2.3* Nuclear Safety Equipment and Cable Location.3.5.2.5 Circuit Failures Due to Common Enclosure Concerns The common enclosure associated circuit concern deals with the possibility of causing Aligns secondary failures due to fire damage to a circuit either whose isolation device fails to isolate the cable fault or protect the faulted cable from reaching its ignition temperature, or the fire somehow propagates along the cable into adjoining fire areas.The electrical circuit design for most plants provides proper circuit protection in the form of circuit breakers, fuses and other devices that are designed to isolate cable faults before ignition temperature is reached. Adequate electrical circuit protection and cable sizing are included as part of the original plant electrical design maintained as part of the design change process. Proper protection can be verified by review of as-built drawings and change documentation.

Review the fire rated barrier and penetration designs that preclude the propagation of fire from one fire area to the next to demonstrate that adequate measures are in place to alleviate fire propagation concerns.See Circuit analysis Project Instruction PI 4.4 and Technical Report TR07800-009 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-36 Attachment B NFPA 805 Section 2.4.2.4 Fire Area Assessment 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. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic)].(See Appendix B for considerations when performing the fire area assessments.)

NEI 00-01 Section NEI 00-01 Section Description NFPA 805 Alignment NFPA 805 Reference 3.4 Fire Area Assessment and Compliance Strategies 3.4.1 Criteria/Assumptions 3.4.1.1 [Number of Postulated Fires]3.4.1.2 [Damage to Unprotected Equipment and Cables]3.4.1.3 [Assess Impacts to Required Components]

3.4.1.4 [Manual Actions]By determining the location of each component and cable by fire area and using the cable to equipment relationships described above, the affected safe shutdown equipment in each fire area can be determined.

Using the list of affected equipment in each fire area, the impacts to safe shutdown systems, paths and functions can be determined.

Based on an assessment of the number and types of these impacts, the required safe shutdown path for each fire area can be determined.

The specific impacts to the selected safe shutdown path can be evaluated using the circuit analysis and evaluation criteria contained in Section 3.5 of this document.

Having identified all impacts to the required safe shutdown path in a particular fire area, this section provides guidance on the techniques available for individually mitigating the effects of each of the potential impacts.The following criteria and assumptions apply when performing fire area compliance assessment to mitigate the consequences of the circuit failures identified in the previous sections for the required safe shutdown path in each fire area.Assume only one fire in any single fire area at a time.Assume that the fire may affect all unprotected cables and equipment within the fire area. This assumes that neither the fire size nor the fire intensity is known. This is conservative and bounds the exposure fire that is required by the regulation.

Address all cable and equipment impacts affecting the required safe shutdown path in the fire area. All potential impacts within the fire area must be addressed.

The focus of this section is to determine and assess the potential impacts to the required safe shutdown path selected for achieving post-fire safe shutdown and to assure that the required safe shutdown path for a given fire area is properly protected.

Use manual actions where appropriate to achieve and maintain post fire safe shutdown conditions in accordance with NRC requirements.

Aligns See Nuclear Safety Capability Assessment Technical Report TR08620-312 Introductory section, alignment identified in subsections Aligns Aligns Aligns See Nuclear Safety Capability Assessment Technical Report TR08620-312 See Nuclear Safety Capability Assessment Technical Report TR08620-312 See NSCA ARC software model and TR08620-312 Aligns Aligns See Nuclear Safety Capability Assessment Technical Report TR08620-312 See Nuclear Safety Capability Assessment Technical Report TR08620-312 3.4.1.5 [Cold Shutdown Where appropriate to achieve and maintain cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, use repairs Repairs] to equipment required in support of post-fire shutdown.NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-37

~Q.Attachment B NEPA 805 Section 2.4.2.4 Fire Area Assessment 3.4.1.6 [Assess Compliance With Deterministic Criteria]3.4.1.7 [Consider Additional Equipment]

3.4.1,8 [Consider Instrument Tubing Effects]3.4.2 Methodology for Fire Area Assessment 3.4.2,1 Identify the Affected Equipment by Fire Area Appendix R compliance requires that one train of systems necessary to achieve and maintain hot shutdown conditions from either the control room or emergency control station(s) is free of fire damage (lIl.G.l.a).

When cables or equipment, including associated circuits, are within the same fire area outside primary containment and separation does not already exist, provide one of the following means of separation for the required safe shutdown path(s): -Separation of cables and equipment and associated nonsafety circuits of redundant trains within the same fire area by a fire barrier having a 3-hour rating (llI.G.2.a);

-Separation of cables and equipment and associated nonsafety circuits of redundant trains within the same fire area by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area (lll.G.2.b).;

-Enclosure of cable and equipment and associated non-safety circuits of one redundant train within a fire area in a fire barrier having a one-hour rating. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area (llI.G.2.c).

For fire areas inside noninerted containments, the following additional options are also available:

-Separation of cables and equipment and associated nonsafety circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards (lll.G.2.d);

Installation of fire detectors and an automatic fire suppression system in the fire area (lll.G.2.e);

or -Separation of cables and equipment and associated non-safety circuits of redundant trains by a noncombustible radiant energy shield (llI.G.2.f).

Use exemptions, deviations and licensing change processes to satisfy the requirements mentioned above and to demonstrate equivalency depending upon the plant's license requirements.

Consider selecting other equipment that can perform the same safe shutdown function as the impacted equipment.

In addressing this situation, each equipment impact, including spurious operations, is to be addressed in accordance with regulatory requirements and the NPP's current licensing basis.Consider the effects of the fire on the density of the fluid in instrument tubing and any subsequent effects on instrument readings or signals associated with the protected safe shutdown path in evaluating post fire safe shutdown capability.

This can be done systematically or via procedures such as Emergency Operating Procedures.

Refer to Figure 3-5 for a flowchart illustrating the various steps involved in performing a fire area assessment.

Use the following methodology to assess the impact to safe shutdown and demonstrate Appendix R compliance:

[Refer to hard copy of NEI 00-01 for Figure 3-5]Identify the safe shutdown cables, equipment and systems located in each fire area that may be potentially damaged by the fire. Provide this information in a report format. The report may be sorted by fire area and by system in order to understand the impact to each safe shutdown path within each fire area (see Attachment 5 for an example of an Affected Equipment Report).Aligns See Nuclear Safety Capability Assessment Technical Report TR08620-312 Aligns Aligns See Nuclear Safety Capability Assessment Technical Report TR08620-312 Technical Report TR08620-019.

See ARC software model for impacts.Introductory section, alignment identified in subsections Aligns See NSCA ARC software model and TR08620-312 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-38

.~Attachment B NFPA 805 Section 2.4.2.4 Fire Area Assessment 3.4.2.2 Determine the Shutdown Paths Least Impacted By a Fire in Each Fire Area 3.4.2.3 Determine Safe Shutdown Equipment Impacts 3.4.2.4 Develop a Compliance Strategy or Disposition to Mitigate the Effects Due to Fire Damage to Each Required Component or Cable Based on a review of the systems, equipment and cables within each fire area, determine which shutdown paths are either unaffected or least impacted by a postulated fire within the fire area. Typically, the safe shutdown path with the least number of cables and equipment in the fire area would be selected as the required safe shutdown path. Consider the circuit failure criteria and the possible mitigating strategies, however, in selecting the required safe shutdown path in a particular fire area. Review support systems as a part of this assessment since their availability will be important to the ability to achieve and maintain safe shutdown.

For example, impacts to the electric power distribution system for a particular safe shutdown path could present a major impediment to using a particular path for safe shutdown.

By identifying this early in the assessment process, an unnecessary amount of time is not spent assessing impacts to the frontline systems that will require this power to support their operation.

Based on an assessment as described above, designate the required safe shutdown path(s) for the fire area. Identify all equipment not in the safe shutdown path whose spurious operation or mal-operation could affect the shutdown function.

Include these cables in the shutdown function list. For each of the safe shutdown cables (located in the fire area) that are part of the required safe shutdown path in the fire area, perform an evaluation to determine the impact of a fire-induced cable failure on the corresponding safe shutdown equipment and, ultimately, on the required safe shutdown path. When evaluating the safe shutdown mode for a particular piece of equipment, it is important to consider the equipment's position for the specific safe shutdown scenario for the full duration of the shutdown scenario.

It is possible for a piece of equipment to be in two different states depending on the shutdown scenario or the stage of shutdown within a particular shutdown scenario.

Document information related to the normal and shutdown positions of equipment on the safe shutdown equipment list.Using the circuit analysis and evaluation criteria contained in Section 3.5 of this document, determine the equipment that can impact safe shutdown and that can potentially be impacted by a fire in the fire area, and what those possible impacts are.The available deterministic methods for mitigating the effects of circuit failures are summarized as follows (see Figure 1-2): -Provide a qualified 3-fire rated barrier. -Provide a 1-hour fire rated barrier with automatic suppression and detection.

-Provide separation of 20 feet or greater with automatic suppression and detection and demonstrate that there are no intervening combustibles within the 20 foot separation distance.

-Reroute or relocate the circuit/equipment, or perform other modifications to resolve vulnerability.

-Provide a procedural action in accordance with regulatory requirements.

-Perform a cold shutdown repair in accordance with regulatory requirements.

-Identify other equipment not affected by the fire capable of performing the same safe shutdown function.

-Develop exemptions, deviations, Generic Letter 86-10 evaluation or fire protection design change evaluations with a licensing change process. Additional options are available for non-inerted containments as described in 10 CFR 50 Appendix R section IIl.G.2.d, e and f. [Refer to hard copy of NEI 00-01 for Figure 1-2]Aligns See NSCA ARC software model and TR08620-312 Aligns Aligns See NSCA ARC software model and TR08620-312 See NSCA ARC software model, results documented in Nuclear Safety Capability Assessment Technical Report TR08620-312 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-39 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-39 ft---Attachment B NFPA 805 Section 2.4.2.4 Fire Area Assessment 3.4.2.5 Document the Compliance Strategy or Disposition Determined to Mitigate the Effects Due to Fire Damage to Each Required Component or Cable Assign compliance strategy statements or codes to components or cables to identify Aligns the justification or mitigating actions proposed for achieving safe shutdown.

The justification should address the cumulative effect of the actions relied upon by the licensee to mitigate a fire in the area. Provide each piece of safe shutdown equipment, equipment not in the path whose spurious operation or mal-operation could affect safe shutdown, and/or cable for the required safe shutdown path with a specific compliance strategy or disposition.

Refer to Attachment 6 for an example of a Fire Area Assessment Report documenting each cable disposition.

[Refer to hard copy of NEI 00-01 for Attachment 6]See NSCA ARC software model, results documented in Nuclear Safety Capability Assessment Technical Report TR08620-312 NEI 04-02 Table B-2 Nuclear Safety Capability Assessment

-Methodology Review Page B-40

.ýý RC-11-0149 Attachment C C. NEI 04-02 Table B-3 Fire Area Transition 471 Pages Attached NEI 04-02 Table B-3 Fire Area Transition Page C-I NEI 04-02 Table B-3 Fire Area Transition Page C-1 3 ° RC-11-0149 Attachment C Fire Area Transition The NSCA provides an evaluation for each fire area at VCSNS to verify that the nuclear safety goals are met. The nuclear safety performance criteria include: Reactivity Control, Inventory and Pressure Control, Decay Heat Removal, Instrumentation, and Support System operability.

The overall NFPA 805 analysis process includes:

1.Identifying equipment needed for safe shutdown, and relevant support equipment, 2.Identifying the cables for this equipment, 3. Locating the fire areas for the cables, 4.Evaluation of the effects of postulated fires in a given fire area, and 5. Resolving open items to ensure the capability to reach and maintain safe shutdown.

The NSCA covers items 4 and 5, evaluation and resolution of safe shutdown capability for each fire area.As previously described, the basic analysis consists of the following steps: 1.Identification of systems and equipment needed to achieve the nuclear safety performance criteria, 2. Selection of cables needed to achieve the nuclear safety criteria, 3. Location and routing of these cables, and 4. Assessment of the ability to achieve the nuclear safety criteria given a fire in each fire area. Steps 1 through 3 are the NSEL, circuit analysis, and routing information described in VCSNS Technical Report TR08620-015, "Nuclear Safety Equipment Report," Project Instruction 4.4,"Cable Selection and Circuit Analysis," Technical Report TR07800-009, "NFPA 805 and Fire PRA Circuit Analysis," and the PC-CKS database.

The analysis of the fire areas (item 4) utilizes the computer codes CAFTA and ARC.The Fire Area analysis may be subdivided into zones and subzones to facilitate the analysis of the ability to achieve safe shutdown.

In particular, where train separation (20 ft horizontal separation plus suppression and detection) was credited, the fire area was divided into zones (and if necessary, subzones).

Controls to preclude transient combustibles and engineering analyses support the zone and subzone designations.

For this evaluation several zones/subzones were combined in the Auxiliary Building to create eight (8) Scenario Areas separated by 3-hour walls. The new Scenario Areas entailed a review of the 20 foot separation for the combined zones to assure that separation was retained.

The Table B-3 for the Auxiliary Building includes these Scenario Area descriptions.

There were also zone/subzone combinations analyzed for the Reactor Building and part of the Intermediate Building.

Again, the purpose of the zone/subzone designation is to assure that 20 foot physical separation is provided to help assure Train separation.

The CAFTA provides a logic model of the plant equipment.

The CAFTA tree is set up to determine whether the safety goals are satisfied, and connects the functional states of the equipment in the NSEL in a way that represents the plant configuration.

In general, there are multiple paths available to achieve a goal, the CAFTA tree connects these paths as the piping and electrical systems are configured.

The ARC Software identifies the equipment and cables in a given fire area and uses the CAFTA model to check whether the safety goal is achieved.

If the goals identified on the ARC button panel are achieved, the button is green, if not, the button is red. The program allows the application of "corrective actions" to achieve success, which are maintained and tabulated for each type and fire area.NEI 04-02 Table B-3 Fire Area Transition Page C-2 RC-11-0149 Attachment C Identification and resolution of open items is integral with the analysis process. The corrective actions needed to achieve the safety goals may include cable rerouting, Electrical Raceway Fire Barrier Systems (ERFBS), equipment modification, and other techniques.

These are identified individually in ARC. The information is exported from ARC and given a unique DROID number. DROIDs that are resolved deterministically retain their DROID tracking and are also tracked with an appropriate process, such as an ECR number. DROIDs that are not amenable to a deterministic resolution are identified with a VFDR number and are described in the B-3 Table. Finally, the input information needed to perform the NSCA, and the salient output and resolution tracking are entered into the FSA database.

The FSA database provides information that will be reviewed for future plant modifications.

This will help assure that equipment and assumptions associated with this analysis are maintained and appropriately considered.

NEI 04-02 Table B-3 Fire Area Transition Page C-3 NEI 04-02 Table B-3 Fire Area Transition Page C-3 SWE " RC-11-0149 Attachment C NEI 04-02 Table B-3 Fire Area Transition Page C-4 NEI 04-02 Table B-3 Fire Area Transition Page C-4 RC-11-0149 Attachment C---U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U -U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U U mm U U-m NEI 04-02 Table B-3 Fire Area Transition Page C-5 NEI 04-02 Table B-3 Fire Area Transition Page C-5

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-.A RC-11-0149 Attachment C I NEI 04-02 Table B-3 Fire Area Transition Page C-63 NEI 04-02 Table B-3 Fire Area Transition Page C-63 RC-11-0149 Attachment C--I ---I I I -* -I -* -I -I -* -I I I -I I I -NEI 04-02 Table B-3 Fire Area Transition Page C-64 NEI 04-02 Table B-3 Fire Area Transition Page C-64 RC-11-0149 Attach ment C--I I--I --NEI 04-02 Table B-3 Fire Area Transition Page C-65 A -,_e- RC-11-0149 Attachment C* -I -I -I ---I I-NEI 04-02 Table B-3 Fire Area Transition Page C-66 NEI 04-02 Table B-3 Fire Area Transition Page C-66 RC-11-0149 Attachment C--I -I -I I I I I I I I I I I I I I I I I I NEI 04-02 Table 8-3 Fire Area Transition Page C-67 NEI 04-02 Table B-3 Fire Area Transition Page C-67

O.f RC-11-0149 Attachment C-=-NEI 04-02 Table 6-3 Fire Area Transition Page C-68 NEI 04-02 Table B-3 Fire Area Transition Page C-68 RC-11-0149 Attach iment C NEI 04-02 Table B-3 Fire Area Transition Page C-69 NEI 04-02 Table B-3 Fire Area Transition Page C-69 RC-11-0149 Attachment C I I--I I I I I I NEI 04-02 Table B-3 Fire Area Transition Page C-70 NEI 04-02 Table B-3 Fire Area Transition Page C-70

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-VE&-G- RC-11-0149 Attachment C i--I I I I I I NEI 04-02 Table B-3 Fire Area Transition Page C-73 NEI 04-02 Table B-3 Fire Area Transition Page C-73 RC-11-0149 Attachment C--I -I -I I I I I I I I NEt 04-02 Table B-3 Fire Area Transition Page C-74 NEI 04-02 Table B-3 Fire Area Transition Page C-74 A ýt-G RC-11-0149 Attachment C--I I--I-U-U-m M-NEI 04-02 Table B-3 Fire Area Transition Page C-75 NEI 04-02 Table B-3 Fire Area Transition Page C-75

.ý-d~ RC-11-0149 Attachment C--I I I I--* -* -I -NEI 04-02 Table B-3 Fire Area Transition Page C-76 NEI 04-02 Table B-3 Fire Area Transition Page C-76 RC-11-0149 Attachment C--I -I -M--I NEI 04-02 Table B-3 Fire Area Transition Page C-77 NEI 04-02 Table B-3 Fire Area Transition Page C-77 RC-11-0149 Attachment C--I -I -I -I -I I NEI 04-02 Table B-3 Fire Area Transition Page C-78 NEI 04-02 Table B-3 Fire Area Transition Page C-78 RC-11-0149 Attachment C--I -I -I -I -I I I I I I I I I I I I NE 40 al -ieAraTasto aeC7 NEI 04-02 Table B-3 Fire Area Transition Page C-79 RC-11-0149 Attachment C--I -I -NEI 04-02 Table B-3 Fire Area Transition Page C-80 NEI 04-02 Table B-3 Fire Area Transition Page C-80 RC-11-0149 Attachment C--I -I --I I I I I I I I I -* -I -I -* -I -* -I -I -NEI 04-02 Table B-3 Fire Area Transition Page C-81 NEI 04-02 Table B-3 Fire Area Transition Page C-81

-9 ,--G-- RC-11-0149 Attachment C I -I I -I -I I--NEI 04-02 Table B-3 Fire Area Transition Page C-82 NEI 04-02 Table B-3 Fire Area Transition Page C-82 RC-11-0149 Attachment C I--I -I -I -I -NEI 04-02 Table B-3 Fire Area Transition Page C-83 NEI 04-02 Table B-3 Fire Area Transition Page C-83 RC-11-0149 Attachment C--I -I -I -* -I -* -* -I -NEI 04-02 Table B-3 Fire Area Transition Page C-84 NEI 04-02 Table B-3 Fire Area Transition Page C-84 RC-11-0 149-Attachment C--I ----I I-U U UýýM-NEI 04-02 Table B-3 Fire Area Transition Page C-85 NEI 04-02 Table B-3 Fire Area Transition Page C-85 RC-11-0149 Attachment C--I I I I I I I I I I I I I NEI 04-02 Table B-3 Fire Area Transition Page C-86 NEI 04-02 Table B-3 Fire Area Transition Page C-86

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RC-11-0149 Attachment C--I II--I I U NI 040ITbeIIIFrIAe Tasiin ae -1 NEI 04-02 Table B-3 Fire Area Transition Page C-1 12 l w RC-11-0149 Attach ment C--I ---_--I-U U U U NEI 04-02 Table B-3 Fire Area Transition Page C-113 NEI 04-02 Table B-3 Fire Area Transition Page C-1 13 W" -- RC-11-0149 Attachment C-------NEI 04-02 Table B-3 Fire Area Transition Page C-114 NEI 04-02 Table B-3 Fire Area Transition Page C-1 14

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Attachment C-* **m m NEI 04-02 Table B-3 Fire Area Transition Page C-120 i RC-11-0149 Attachment C--I-I I--I--I NEI 04-02 Table B-3 Fire Area Transition Page C-121 NEI 04-02 Table B-3 Fire Area Transition Page C-121

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' CL.RC-11-0149 Attachlment C.II II--I ---I I NEI 04-02 Table B-3 Fire Area Transition Page C-133 RC-11-0149 Attachment C I--I I I I NEI 04-02 Table B-3 Fire Area Transition Page C-134 NEI 04-02 Table B-3 Fire Area Transition Page C-134 RC-11-0149 Attachment C NEI 04-02 Table B-3 Fire Area Transition Page C-135 NEI 04-02 Table B-3 Fire Area Transition Page C-135 U.RC-11-0149 Attachment C--i i i--I ---I -NEI 04-02 Table B-3 Fire Area Transition Page C-136 NEI 04-02 Table B-3 Fire Area Transition Page C -136

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RC-11-0149 Attachment C--I I I--I I NEI 04-02 Table B-3 Fire Area Transition Page C-184 NEI 04-02 Table B-3 Fire Area Transition Page C-184 RC-11-0149 Attachment C--I I--I I I NEI 04-02 Table B-3 Fire Area Transition Page C-185 NEI 04-02 Table B-3 Fire Area Transition Page C-185 RC-11-0149 Attachment C--I I I I NEI 04-02 Table B-3 Fire Area Transition Page C-186 NEI 04-02 Table B-3 Fire Area Transition Page C-186 RC-11-0 149 -*Atcmn--I-U U U-U Attachment C 0 -NEI 04-02 Table B-3 Fire Area Transition Page C-187 NEI 04-02 Table B-3 Fire Area Transition Page C-187 RC-11-0149 Attachment C II II--I--I F- I NEI 04-02 Table B-3 Fire Area Transition Page C-188 NEI 04-02 Table B-3 Fire Area Transition Page C-188 RC-11-0149 Attachment C--B NEI 04-02 Table B-3 Fire Area Transition Page C-189 NEI 04-02 Table B-3 Fire Area Transition Page C-189

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RC110149 Attachment C--I I-l--I--NEI 04-02 Table B-3 Fire Area Transition Page C-195 NEI 04-02 Table B-3 Fire Area Transition Page C-195 RC-11-0149 Attachment C--I I--I--I I NEI 04-02 Table B-3 Fire Area Transition Page C-196 RC-11-0149 Attachment C--I--I--I--I NEI 04-02 Table B-3 Fire Area Transition Page C-197 NEI 04-02 Table B-3 Fire Area Transition Page C-197 A- RC-11-0149 Attachment C--I NEI 04-02 Table B-3 Fire Area Transition Page C-198 NEI 04-02 Table B-3 Fire Area Transition Page C-198 RC-11-0149 Attachment C II II--I II I-- I NEI 04-02 Table B-3 Fire Area Transition Page C-199 NEI 04-02 Table B-3 Fire Area Transition Page C-199 SvAJC"'-f-RC-11-0149 Attachment C U--I-U--I U--I-U NEI 04-02 Table B-3 Fire Area Transition Page C-200 NEI 04-02 Table B-3 Fire Area Transition Page C-200

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RC-11-0149 Attachment C--I -I ---I-M II-li-NEI 04-02 Table B-3 Fire Area Transition Page C-205 NEI 04-02 Table B-3 Fire Area Transition Page C-205 A RC-11-0149 Attachment C--I I I I--I I I I I NEI 04-02 Table B-3 Fire Area Transition Page C-206 NEI 04-02 Table B-3 Fire Area Transition Page C-206

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RC-11-0149 Attachment C--NEI 04-02 Table B-3 Fire Area Transition Page C-218 NEI 04-02 Table B-3 Fire Area Transition Page C-218

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RCA 10149 Attachment C--I--I--I--I--I NEI 04-02 Table B-3 Fire Area Transition Page C-219 w A- ft~ a RC-11-0149 Attach ment C NEI 04-02 Table B-3 Fire Area Transition Page C-220 NEI 04-02 Table B-3 Fire Area Transition Page C-220 r-SM -EA--- RC-11-0149 Attachment C--I--I I I--I I I NEI 04-02 Table B-3 Fire Area Transition Page C-221 NEI 04-02 Table B-3 Fire Area Transition Page C-221 RC-11-0149 Attachment C NEI 04-02 Table B-3 Fire Area Transition Page C-222 NEI 04-02 Table B-3 Fire Area Transition Page C-222 RC-11-0149 Attachment C--I I I I--I ---I -NEI 4-0 Tale -3 Fre reaTrasitin Pge -2I NEI 04-02 Table B-3 Fire Area Transition Page C-223

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-RC-11-0149 Attach ment C--I II NEI 04-02 Table B-3 Fire Area Transition Page C-227 NEI 04-02 Table B-3 Fire Area Transition Page C-227

-- RC-11-0149 Attachment C III NEI 04-02 Table B-3 Fire Area Transition Page C-228 NEI 04-02 Table B-3 Fire Area Transition Page C-228 RC-11-0149 Attachment C UI I--I U I--I U NEI 04-02 Table B-3 Fire Area Transition Page C-229 NEI 04-02 Table B-3 Fire Area Transition Page C-229 RC-11-0149 Attachment C---NEI 04-02 Table B-3 Fire Area Transition Page C-230 NEI 04-02 Table B-3 Fire Area Transition Page C-230

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RC-11-0149 Attachment C--I ---_--I -I -I -I ---I NEI 04-02 Table B-3 Fire Area Transition Page C-232 NEI 04-02 Table B-3 Fire Area Transition Page C-232 RC-11-0149 Attachment C U EE-I--I NEI 04-02 Table B-3 Fire Area Transition Page C-233 NEI 04-02 Table B-3 Fire Area Transition Page C-233 v=E--- W.' RC-11-0149 Attachment C NEI 04-02 Table B-3 Fire Area Transition Page C-234 NEI 04-02 Table B-3 Fire Area Transition Page C-234

RC-11-0149 Attachment C-m[ * * *m m NEI 04-02 Table B-3 Fire Area Transition Page C-235 RC-11-0149 Attachment C--I ---I I-I I NEI 04-02 Table B-3 Fire Area Transition Page C-236 NEI 04-02 Table B-3 Fire Area Transition Page C-236 RC-11-0149 Attachment C--B UI NEI 04-02 Table B-3 Fire Area Transition Page C-237 NEI 04-02 Table B-3 Fire Area Transition Page C-237 RC-11-0149 Attachment C--B-m -NEI 04-02 Table B-3 Fire Area Transition Page C-238 NEI 04-02 Table B-3 Fire Area Transition Page C-238

~~jpq~A" -- -.- RC-11-0149 Attachment C NEI 04-02 Table B-3 Fire Area Transition Page C-239 NEI 04-02 Table B-3 Fire Area Transition Page C-239

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RC-11-0149 Attachment C---_ _ _NEI 04-02 Table B-3 Fire Area Transition Page C-246 NEI 04-02 Table B-3 Fire Area Transition Page C-246 RC-11-0149 Attachment C U NEI 04-02 Table B-3 Fire Area Transition Page C-247 NEI 04-02 Table B-3 Fire Area Transition Page C-247

.F-WS-- RC-11-0149 Attachment C---_ _ _ _NEI 04-02 Table 6-3 Fire Area Transition Page C-248 NEI 04-02 Table B-3 Fire Area Transition Page C-248 A *Wý-06~ RC-11-0149 Attachment C II NEI 04-02 Table B-3 Fire Area Transition Page C-249 NEI 04-02 Table B-3 Fire Area Transition Page C-249

-=--a- RC-11-0149 Attachment C-m --_ _ _-n NEI 04-02 Table B-3 Fire Area Transition Page C-250 NEI 04-02 Table B-3 Fire Area Transition Page C-250

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