ML13127A397: Difference between revisions

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-Hot operating condition (Reactor mode switch is in "Shutdown" position and reactor coolant temperature is greater than 212'F), with the Shutdown Cooling (SDC) system not aligned for decay heat removal.* Non-Power analysis for potential fires while in Shutdown Condition  
-Hot operating condition (Reactor mode switch is in "Shutdown" position and reactor coolant temperature is greater than 212'F), with the Shutdown Cooling (SDC) system not aligned for decay heat removal.* Non-Power analysis for potential fires while in Shutdown Condition  
-Hot operating condition and lower operating conditions.
-Hot operating condition and lower operating conditions.
A copy of TS Section 1.1 containing the definitions of the NMP I reactor operating conditions is provided in Figure SSD/CA RAI 01-1 below to facilitate a clear understanding of the analytical coverage provided by the two analyses described above.2 of 47 ENCLOSURE1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)1.0 DEFINITION4S 1.1 Reactor Operatina Conditions The various reactor operating conditions are defined below. Individual technical specifications amplify these definitions when appropriate.
A copy of TS Section 1.1 containing the definitions of the NMP I reactor operating conditions is provided in Figure SSD/CA RAI 01-1 below to facilitate a clear understanding of the analytical coverage provided by the two analyses described above.2 of 47 ENCLOSURE1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)1.0 DEFINITION4S
 
===1.1 Reactor===
Operatina Conditions The various reactor operating conditions are defined below. Individual technical specifications amplify these definitions when appropriate.
a- Shutdown Condition  
a- Shutdown Condition  
-Cold (1) The reactor mode switch is in the shutdown position or refuel position.(2) No core alterations leading to an addition of reactivity are being performed.
-Cold (1) The reactor mode switch is in the shutdown position or refuel position.(2) No core alterations leading to an addition of reactivity are being performed.
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Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements structures housing equipment required for nuclear plant operations are considered as"power block" structures.
Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements structures housing equipment required for nuclear plant operations are considered as"power block" structures.
These structures are listed in Attachment I and define the "power block" and "plant".4.2 Nuclear Safety Performance Criteria The Nuclear Safety Performance Criteria are established in Section 1.5 of NFPA 805.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.
These structures are listed in Attachment I and define the "power block" and "plant".4.2 Nuclear Safety Performance Criteria The Nuclear Safety Performance Criteria are established in Section 1.5 of NFPA 805.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" Establishing Recovery Actions" 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 4.3.4.2.1.1 Compliance with NFPA 805 Section 2.4.2 Overview of Process NFPA 805 Section 2.4.2 Nuclear Safety Capability Assessment states: "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" Establishing Recovery Actions" 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 4.3.4.2.1.1 Compliance with NFPA 805 Section 2.4.2 Overview of Process NFPA 805 Section 2.4.2 Nuclear Safety Capability Assessment states: "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 post-fire safe shutdown analysis (SSA) methodology against the guidance provided in NEI 00-01, Revision 2, Chapter 3,"Deterministic Methodology," as discussed in Appendix B-2 of NEI 04-02. NMP1 used the guidance provided in NEI 00-01, Revision 2 because it is endorsed as an NMI, pil213Pge1 I NMP1, April 2013 Page 14 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements acceptable methodology in NRC RG 1.205 and due to feedback received as a result of NRC requests for additional information on other post-pilot plant LARs.The methodology is depicted in Figure 4-2 and consisted of the following activities: " Each specific subsection of NFPA 805 Section 2.4.2 was correlated to the corresponding section of Chapter 3 of NEI 00-01, Revision 2. 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 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 those sections that do not align, an assessment was made to determine if the failure to maintain strict alignment with the guidance in NEI 00-01 could have adverse consequences.
(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 post-fire safe shutdown analysis (SSA) methodology against the guidance provided in NEI 00-01, Revision 2, Chapter 3,"Deterministic Methodology," as discussed in Appendix B-2 of NEI 04-02. NMP1 used the guidance provided in NEI 00-01, Revision 2 because it is endorsed as an NMI, pil213Pge1 I NMP1, April 2013 Page 14 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements acceptable methodology in NRC RG 1.205 and due to feedback received as a result of NRC requests for additional information on other post-pilot plant LARs.The methodology is depicted in Figure 4-2 and consisted of the following activities: " Each specific subsection of NFPA 805 Section 2.4.2 was correlated to the corresponding section of Chapter 3 of NEI 00-01, Revision 2. 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 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 those sections that do not align, an assessment was made to determine if the failure to maintain strict alignment with the guidance in NEI 00-01 could have adverse consequences.
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-Specific references to Nuclear Safety Capability Assessment Documents." Licensing Actions -Specific references to safety evaluations that will remain part of the post-transition licensing basis. A brief description of the condition and the basis for acceptability of the licensing action are provided.
-Specific references to Nuclear Safety Capability Assessment Documents." Licensing Actions -Specific references to safety evaluations that will remain part of the post-transition licensing basis. A brief description of the condition and the basis for acceptability of the licensing action are provided.
In addition, summaries of Fire Risk Evaluations performed for variances from the deterministic requirements are also provided." EEEE -Specific references to EEEE 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 are provided." VFDRs -Specific variances from the deterministic requirements of NFPA 805 Section 4.2.3. Refer to Section 4.5.2 for a discussion of the performance-based approach.4.3 Non-Power Operational Modes 4.3.1 Overview of Evaluation Process NMP1 implemented the process outlined in NEI 04-02 and FAQ 07-0040, "Non-Power Operations Clarifications." The goal (as depicted in Figure 4-5) is to ensure that contingency plans are established when the plant is in a Non-Power Operational (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: " Review of the existing Outage Management Processes" Identification of Equipment/Cables:
In addition, summaries of Fire Risk Evaluations performed for variances from the deterministic requirements are also provided." EEEE -Specific references to EEEE 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 are provided." VFDRs -Specific variances from the deterministic requirements of NFPA 805 Section 4.2.3. Refer to Section 4.5.2 for a discussion of the performance-based approach.4.3 Non-Power Operational Modes 4.3.1 Overview of Evaluation Process NMP1 implemented the process outlined in NEI 04-02 and FAQ 07-0040, "Non-Power Operations Clarifications." The goal (as depicted in Figure 4-5) is to ensure that contingency plans are established when the plant is in a Non-Power Operational (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: " Review of the existing Outage Management Processes" Identification of Equipment/Cables:
o Review of plant systems to determine success paths that support each of the defense-in-depth Key Safety Functions (KSFs), and o Identification of cables required for the selected components and determination of their routing." Perform Fire Area Assessments (identify pinch points -plant locations where a single fire may damage all success paths of a KSF)." Manage pinch-points associated with fire-induced vulnerabilities during the outage.!NMPI, April 2013 Page 25 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements The process is depicted in Figures 4-5 and 4-6. The results are presented in Section 4.3.2.Figure 4-5 Review POSs, KSFs, Equipment, and Cables, and Identify Pinch Points NMPI, April 2013 Page 26 I I NIVIP1, April 2013 Page 26 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements 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 Figure 4-6 Manage Pinch Points NMPI, April 2013 Page 27 IlNMP1, April 2013 Page 27 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements 4.3.2 Results of the Evaluation Process Based on FAQ 07-0040, the Plant Operating States (POSs) considered for equipment and cable selection are defined in EIR 51-9171174, Nine Mile Point 1, Non-Power Operations KSF Equipment List. The methodology for determination of KSFs, success paths, components required to achieve the success paths and their associated cabling arewere defined in EIR 51-9137629, Nine Mile Point 1 -Nuclear Power Station -NFPA 805 Transition  
o Review of plant systems to determine success paths that support each of the defense-in-depth Key Safety Functions (KSFs), and o Identification of cables required for the selected components and determination of their routing." Perform Fire Area Assessments (identify pinch points -plant locations where a single fire may damage all success paths of a KSF)." Manage pinch-points associated with fire-induced vulnerabilities during the outage.!NMPI, April 2013 Page 25 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements The process is depicted in Figures 4-5 and 4-6. The results are presented in Section 4.3.2.Figure 4-5 Review POSs, KSFs, Equipment, and Cables, and Identify Pinch Points NMPI, April 2013 Page 26 I I NIVIP1, April 2013 Page 26 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements 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 Figure 4-6 Manage Pinch Points NMPI, April 2013 Page 27 IlNMP1, April 2013 Page 27 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements
 
====4.3.2 Results====
of the Evaluation Process Based on FAQ 07-0040, the Plant Operating States (POSs) considered for equipment and cable selection are defined in EIR 51-9171174, Nine Mile Point 1, Non-Power Operations KSF Equipment List. The methodology for determination of KSFs, success paths, components required to achieve the success paths and their associated cabling arewere defined in EIR 51-9137629, Nine Mile Point 1 -Nuclear Power Station -NFPA 805 Transition  
-Non Power Operations Component Pinch Point Analysis.
-Non Power Operations Component Pinch Point Analysis.
These documents provide the component selection information for the POSs included in the analysis (HSD, CSD, Refueling and Defueled conditions).
These documents provide the component selection information for the POSs included in the analysis (HSD, CSD, Refueling and Defueled conditions).

Revision as of 01:13, 13 October 2018

Nine Mile Point, Unit 1, License Amendment Request Pursuant to 10 CFR 50.90: Adoption of NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition) - Response to NRC Rai. Part
ML13127A397
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 04/30/2013
From: Costanzo C R
Constellation Energy Group, EDF Group, Nine Mile Point
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
TAC ME8899
Download: ML13127A397 (238)
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ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Nine Mile Point Nuclear Station, LLC April 30, 2013 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)By letter dated June 11, 2012, Nine Mile Point Nuclear Station, LLC (NMPNS) requested an amendment to the Nine Mile Point Unit 1 (NMP1) Renewed Facility Operating License DPR-63. The proposed amendment would adopt a new risk-informed performance-based (RI-PB) fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a) and 10 CFR 50.48(c);

the guidance in Regulatory Guide (RG) 1.205, "Risk-Informed Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1; and National Fire Protection Association (NFPA) 805,"Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition.This enclosure provides supplemental information in response to the following three requests for additional information (RAIs) documented in the NRC's letter dated January 3, 2013: Safe Shutdown /Circuit Analysis RAI 01, RAI 03, and RAI 07. NMPNS agreed to provide responses to these three RAIs by April 30, 2013. Each individual NRC RAI is repeated (in italics), followed by the NMPNS response.Safe Shutdown / Circuit Analysis RAI 01 The description in LAR Section 4.2.1.2 of safe and stable is defined as, "the ability to maintain Keff < 0. 99 with a reactor coolant temperature at or below the requirement for hot shutdown and then subsequently cool down and maintain NMP1 in a cold shutdown condition. " The nuclear safety capability assessment (NSCA) methodology review (LAR Attachment B) includes discussion of cold shutdown (CSD)methodology as appropriate and the methods for meeting performance goals in the fire area assessments (LAR Attachment C) include CSD components and systems.Additional information is needed regarding the timing, systems, actions, and any repairs, necessary to achieve and maintain CSD. There is no discussion of the risk associated with actions to achieve and maintain CSD.VFDRs are identified in LAR Attachment Cforperfornance criteria related to CSD. In some cases, these VFDRs are dispositioned on the basis that the risk, defense-in-depth (DID), and safety margins meet the acceptance criteria of NFPA 805 with a recovery action (RA) credited.

The VFDR disposition further states the RA has been evaluated for feasibility and reliability within the FPRA using HRA methods (e.g., Attachment C, pg. 64, VFDR-05-025).

Additional information is needed to address the following specific issues: a. Provide the timing assumed for sustaining hot shutdown (once achieved) and then transitioning from hot shutdown to, and achieving CSD.b. Describe how cold shutdown was modeled in the FPRA, including the risk of RAs credited for disposition of VFDRs associated with CSD NSCA equipment.

c. System or component capacity limitations are not specifically described for each applicable performance goal. Provide a description of capacity limitations, need to replenish systems, and time-critical actions for other systems needed to maintain safe and stable conditions (e.g., nitrogen supply for valve operations, water supplies, boron supply, DC battery power, fuel, etc.).1 of 47 ENCLOSURE1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)d. Describe in more detail the resource (staffing) requirements, timing, and feasibility of operator actions to recover NSCA equipment to achieve and sustain safe and stable conditions.
e. Attachment G describes actions involving repairs to valve and pump wiring for shutdown cooling. Describe in more detail the resource (staffing) requirements, timing, and feasibility of actions to repair NSCA equipment to achieve and maintain CSD safe and stable conditions.

f Provide a more detailed description of the risk of failure of operator actions and equipment necessary to sustain safe and stable conditions.

g. Describe the actions that are planned for MSOs for shutdown cooling or any time the need to restore decay heat removal is short based on time to boil.Response to Safe Shutdown / Circuit Analysis RAI 01 General Nine Mile Point Nuclear Station, LLC (NMPNS) has elected to modify its NFPA 805 transition analysis for NMP1 to revise the approach for demonstrating the ability to reach and maintain safe and stable conditions, as specified by NFPA 805. The original Nuclear Safety Capability Assessment (NSCA)established as its basis for demonstrating safe and stable conditions the requirement to maintain K~f <0.99 with a reactor coolant temperature at or below the requirements for hot shutdown and then subsequently cool down and maintain the plant in a cold shutdown condition.

Consistent with NFPA 805 and supplemental guidance, NMPNS is revising its basis for the NMP1 NSCA to include only the requirement to establish hot shutdown conditions, including long-term hot shutdown capability.

The response to this RAI, including Parts a through g, is provided within the context of the aforementioned change.Demonstration of the nuclear safety performance criteria for safe and stable conditions is performed in two analyses based on the plant operating modes, as defined in the NMP1 Technical Specifications (TS).These analyses are defined as follows:* At-Power analysis for potential fires while in either: (i) the Power Operating Condition (Reactor mode switch is in "Startup" or "Run" position and the reactor is critical or criticality is possible due to control rod withdrawal), or (ii) the Shutdown Condition

-Hot operating condition (Reactor mode switch is in "Shutdown" position and reactor coolant temperature is greater than 212'F), with the Shutdown Cooling (SDC) system not aligned for decay heat removal.* Non-Power analysis for potential fires while in Shutdown Condition

-Hot operating condition and lower operating conditions.

A copy of TS Section 1.1 containing the definitions of the NMP I reactor operating conditions is provided in Figure SSD/CA RAI 01-1 below to facilitate a clear understanding of the analytical coverage provided by the two analyses described above.2 of 47 ENCLOSURE1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)1.0 DEFINITION4S

1.1 Reactor

Operatina Conditions The various reactor operating conditions are defined below. Individual technical specifications amplify these definitions when appropriate.

a- Shutdown Condition

-Cold (1) The reactor mode switch is in the shutdown position or refuel position.(2) No core alterations leading to an addition of reactivity are being performed.

(3) Reactor coolant temperature is less than or equal to 2121F.b. Shutdown Condition

-Hot (1) The reactor mode switch is in the shutdown position.

    • (2) No core alterations leading to an addition of reactivity are being performed.

(3) Reactor coolant temperature is greater than 212'F.c. Refueling Condition (1) The reactor mode switch is in the refuel position.(2) The reactor coolant temperature is less than 212°F.(3) Fuel may be loaded or unloaded.(4) No more than one operable control rod may be withdrawn.

d. Prwer Operating Condition (1) Reactor mode switch is in startup or run position.(2) Reactor is critical or criticality is possible due to control rod withdrawal.
e. Major Maintenance Condition (1) No fuel is in the reactor.The reactor mode switch may be placed in the startup position to perform the shutdown margin demonstration.

See Special Test Exception 3.7.1.The reactor mode switch may be placed in the refuel position to perform reactor coolant system pressure testing, control rod scram time testing and scram recovery operations.

Figure SSD/CA RAI 01-1: NMP1 Technical Specification Section 1.1, Definitions for Reactor Operating Conditions The practical manifestation of the redefined basis for safe and stable is that the At-Power analysis now includes only equipment necessary to achieve and maintain hot shutdown conditions, including some new equipment required to demonstrate long-term hot shutdown capability.

The NSCA no longer requires the ability to achieve and maintain cold shutdown.

On this basis, equipment associated with the SDC system and any associated variances from deterministic requirements (VFDRs) of NFPA 805 Section 4.2.3 have been removed from the NSCA. Cold shutdown issues are now addressed only within the context of the Non-Power Operations (NPO) analysis, and only to the extent that they apply (refer to the response for Safe Shutdown / Circuit Analysis RAI 03). A formal screening process based on the criteria shown in Figure S SD/CA RAI 01-2 was used to screen and identify VFDRs associated only with the SDC system;i.e., cold shutdown only VFDRs.3 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)1 If the credited HSD path is "C" or"D," as defined in EIR 51-9133191 (NMP1 NSCA), all VFDRs relevant to the fire area are required for long term "Safe and Stable" operation.

Shutdown paths "C" and "D" utilize decay heat removal capabilities independent of the shutdown cooling system, and require the use of Core Spray (CS), Emergency Relief Valves (ERVs), and Containment Spray (CTS)/Containment Spray Raw Water (CTSRW) systems to assure appropriate HSD conditions.

2. The following criteria apply to the screening of VFDRs where the pertinent fire area HSD path is "A" or "B," as defined in EIR 51-9133191 (NMP1 NSCA). Shutdown paths "A" and "B" require the use of Emergency Condenser Cooling (supports HSD) and Shutdown Cooling (supports CSD). Thus, VFDRs may be removed if they are associated with only CSD operation." VFDRs associated with systems or components required for initial plant inventory or pressure control are required for the "At-Power analysis." These systems include Emergency Condenser Cooling (EC), Main Feedwater Isolation, spurious ERV actuation, and Reactor Water Cleanup (RWCU) isolation.

Note that the Control Rod Drive (CRD) System is required to ensure adequate Reactor Pressure Vessel (RPV) makeup is available to account for nominal inventory losses over time, thus VFDRS associated with availability of the system are required for the "At-power analysis." VFDRs associated with systems or components necessary to support vital plant diagnostic indication are required for the "At-Power analysis." These systems include Reactor Coolant System (RCS) pressure and level indication and torus level indication.

Torus level indication is required because increasing torus level may reduce the available margin to transition to CSD and require operators to depressurize earlier in the event, thereby potentially jeopardizing the ability to maintain "stable" conditions while in HSD. Thus, scenarios that could impact the availability of torus level are required for the "At-Power analysis," such as spurious CTSRW Injection)." VFDRs associated with systems or components associated with Reactor Building Closed Loop Cooling, Emergency SW, or Normal SW availability are required for the "At-Power analysis." These systems are necessary to provide control room cooling, and thus must remain available prior to the transition to CSD." VFDRs associated with electrical power availability are generally required for the "At-Power analysis" (e.g. Loss of Train A/B battery charging capability).

Loss of individual power supplies are considered on a case-by-case basis, and are binned in accordance with the function of the equipment supported by the power supply." VFDRs associated with the availability of HVAC units are generally required for the "At-Power analysis" unless the affected equipment/component supports a purely CSD function." VFDRs associated with availability of the Shutdown Cooling System are required for CSD/NPO.Similarly VFDRs associated with Torus cooling are required for CSD/NPO. Torus cooling is not required for the "At-Power analysis" because EC cooling is the primary method for decay heat removal, and the system is not adversely impacted by a loss of torus cooling.Figure SSD/CA RAI 01-2: Criteria for Screening and Identifying Cold Shutdown VFDRs 4 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Changes to the NSCA and License Amendment Request (LAR) documentation necessary to implement the newly established safe and stable analysis basis include:* Elimination of SDC system components from the NSCA equipment list. This change resulted in elimination of 73 VFDRs associated with cold shutdown (see Table SSD/CA RAI 01-1 for a list of removed cold shutdown VFDRs).* Include in the NSCA the fire water system valves associated with refilling the Emergency Condenser makeup tanks to support long-term decay heat removal capability for hot shutdown operation.

This change resulted in the addition of 22 new VFDRs (see Table SSD/CA RAI 01-2 for a list of new VFDRs).* Addition of two new recovery actions associated with manual valve alignment and operation of the DFP to refill the Emergency Condenser makeup tanks to support long-term operation of the Emergency Cooling (EC) system, to satisfy decay heat removal requirements for hot shutdown." Re-quantification of the Fire PRA model and re-calculation of ACDF and ALERF. In both cases, slight improvements in risk were realized as a result of the changes (refer to LAR Attachments C and W for specific ACDF and ALERF values). Although delta risk decreased overall, there were slight increases in the contributions from recovery actions for ACDF and ALERF. This small increase is attributable to the reduction in reliance on SDC system components and increased reliance on the DFP and fire system valves for demonstrating that safe and stable conditions are achieved and maintained.

The net effect makes the post-transition plant more like the deterministically compliant plant in terms of risk. The new approach to demonstrating safe and stable conditions results in a slight increase in reliance on recovery actions due to the addition of the two new recovery actions noted above.* Updates to the following LAR Transition Report sections and attachments:

-Update Section 4.2 to incorporate the new basis for safe and stable, including discussion on long-term maintenance of hot shutdown conditions (see Enclosures 3 and 4).-Update Table 4-3 to capture summary-level changes to the analysis (see Enclosures 3 and 4).-Update Attachment A (Table B-i), Section 3.5.16, to address the new time frame for alternate use of the DFP to refill the Emergency Condenser makeup tanks (see Enclosures 3 and 4).-Update Attachment B (Table B-2) to address the revised methodology for achieving and maintaining safe and stable conditions (see Enclosures 3 and 4).-Update Attachment C (Table B-3) to remove cold shutdown VFDRs, add new VFDRs associated with the DFP, and update the fire risk summary results (see Enclosures 5 and 6).-Update Attachment G to add new recovery actions associated with manual alignment and operation of the DFP to refill the Emergency Condenser makeup tanks (see Enclosures 3 and 4).5 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)-Update the Attachment W Fire PRA insights and results to reflect re-quantification of the Fire PRA model and re-calculation of ACDF and ALERF values with the VFDR changes considered (see Enclosures 5 and 6).Specific responses to Parts a through g of this RAI are provided below and are based on the revised safe and stable analysis basis. Thus, in some cases, the questions pertaining only to cold shutdown are no longer relevant.Part a Sustaining hot shutdown conditions (once achieved) for an extended period of time is accomplished by (1) ensuring a continual source of water to the Emergency Condensers in support of decay heat removal using the EC system, (2) ensuring a long-term source of inventory for makeup to the reactor, and (3) ensuring continual operation of at least one emergency diesel generator to supply AC power to the electrical distribution system.Upon achieving hot shutdown conditions, the plant is able to maintain safe and stable operation for an extended period of time using the EC system. After 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the Emergency Condenser makeup tanks can be replenished as needed using the DFP, which draws water from Lake Ontario (effectively an infinite source). In the event water from the condensate storage tanks (CST) can be transferred to the Emergency Condenser makeup tanks, operation of the DFP would not be required until some point beyond 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Periodic refueling of the DFP is accomplished in accordance with existing plant procedures using the DFP fuel oil storage tank. The DFP day tank contains sufficient fuel for 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> of operation.

The DFP fuel oil storage tank contains fuel to support 6.1 days of operation.

Reactor coolant makeup is required after 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, assuming a nominal TS leakage rate of 25 gpm.Makeup is provided via the Control Rod Drive (CRD) system using one of the CRD pumps drawing suction from the CST. Alternating current (AC) power is required to operate a CRD pump. The DFP may be aligned to provide reactor coolant makeup in the event that no CRD pump is available.

In the event the EC system is not available, the plant can be maintained in hot shutdown by opening three Electromatic Relief Valves (ERVs) in the automatic depressurization system (ADS) and blowing steam to the Torus to reduce pressure.

When reactor pressure reaches approximately 365 psig, the Core Spray (CS) system may be utilized to provide core cooling. Both AC and direct current (DC) electrical power are required for this method of decay heat removal. This alternate means of decay heat removal can be used to maintain safe and stable conditions until such time that the SDC system is placed in service. The CS system is a two loop system. Operation of one loop is adequate to ensure core cooling. When utilizing the CS system, the ERVs pass steam and then, eventually, water to the Torus to remove decay heat from the reactor, in essence placing the Reactor Coolant System (RCS) in recirculation through the Torus. During this process, decay heat is removed by operation of the Containment Spray (CTS) system in conjunction with the Containment Spray Raw Water (CTSRW) system. This method of decay heat removal negates the need for another system to provide inventory makeup. AC power is required to initiate and maintain this method of decay heat removal;thus, long-term maintenance of this operating mode is dependent on maintaining AC electrical power.For either of the hot shutdown methods used to achieve and maintain long-term safe and stable conditions, AC power availability from either the station Emergency Diesel Generators (EDGs) or offsite power is necessary.

Offsite power is not credited in the NSCA. The EDGs can be refueled in 6 of 47 ENCLOSURE1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)accordance with existing plant procedures using an on-site fuel source (tanker truck), until such time that offsite power is restored.

Refueling of a continually operating EDG is estimated to be required after four days (assuming one EDG operating at full load). Given the long timeframe before EDG and/or DFP refueling is necessary, additional resources from the emergency response organization will be available to support EDG and DFP refueling activities.

Transition to cold shutdown is no longer a requirement in the NSCA for ensuring that safe and stable conditions are achieved and maintained.

Part b Success in the Fire PRA is defined to be a controlled stable state with the reactor subcritical, its water inventory stable, and its heat being removed. The Fire PRA success criteria require that this stable condition be maintained for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; i.e., cold shutdown is not required for success in the Fire PRA.Nevertheless, some equipment that may be used to establish cold shutdown conditions is modeled in the Fire PRA; e.g., the SDC system is modeled in the Fire PRA to provide a backup to the loss of other heat removal systems.The definition of safe and stable conditions for at-power analysis in LAR Section 4.2.1.2 is now revised to require achieving and maintaining hot shutdown conditions for an extended period of time, rather than to require achieving hot shutdown conditions and transitioning to cold shutdown.

Some of the VFDRs originally presented in the LAR were identified as VFDRs only because they presented a challenge in meeting the nuclear safety performance criteria associated with achieving cold shutdown.Due to the elimination of the requirement for cold shutdown from the definition of safe and stable conditions, the cold-shutdown VFDRs have been eliminated (see listing in Table SSD/CA RAI 01-1).Thus, it is no longer necessary to evaluate the change in risk (ACDF and ALERF), including the risk of recovery actions, associated with cold-shutdown VFDRs.Part c The NMP1 NSCA (EIR 51-9133191) was developed in accordance with the NFPA 805 requirements and applicable Frequently Asked Questions (FAQs). Section 1.5.1 of NFPA 805 identifies the pertinent nuclear safety performance criteria that are to be satisfied in order to "provide reasonable assurance that, in the event of a fire, the plant is not placed in an unrecoverable condition." The criteria are:* Reactivity Control* Inventory and Pressure Control* Decay Heat Removal" Vital Auxiliaries

  • Process Monitoring LAR Attachment C (Table B-3) documents how each performance criteria is satisfied on a fire area basis. When applicable, VFDRs are identified for each performance goal in Table B-3 and a disposition is provided.

The revised NMP1 basis for safe and stable requires that hot shutdown conditions be achieved and maintained for an extended period of time, which introduces additional 7 of 47 ENCLOSURE1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)system capacity limitations to ensure that the nuclear safety performance criteria are satisfied.

The additional capacity limitations include:* A source of inventory makeup to the Emergency Condensers to ensure availability of the EC system for extended hot shutdown operation.

  • A source of inventory makeup to the reactor to compensate for primary system leakage over an extended time frame. Reactor inventory makeup was addressed originally as a cold shutdown consideration, whereas the revised NMPI basis for safe and stable could require that inventory makeup be established prior to the cold shutdown transition.
  • A source of fuel oil to assure long-term availability of the EDGs and DFP.The response to Part a of this RAI provides details for these three identified long-term considerations.

Additional VFDRs (applicable to the pertinent fire areas) have been added to LAR Attachment C (Table B-3) to address deterministic concerns regarding availability of the Fire Water System.Specifically, local start of the DFP is addressed by plant procedure N1-OP-21A.

The procedure also address actions necessary to replenish fuel oil to the DFP, thereby ensuring adequate fuel oil supply, and other actions necessary to ensure pump operation if instrument air is not available.

Procedure N1-PM-V 19 provides guidance to replenish the EDG fuel oil storage tank to assure continued long-term operation of the EDGs.Time-critical actions needed to support capacity limitations involve load shedding for Battery Boards 11 and 12 to ensure DC power availability.

These actions are included in Table G-1 of LAR Attachment G.Part d Recovery actions credited in the NFPA 805 transition to bring the plant to and maintain it in a safe and stable condition (i.e., hot shutdown) fall into one of two categories, as follows: " Recovery actions modeled in the Fire PRA and analyzed as part of the human reliability analysis (HRA) of the Fire PRA. Appendix I of the Human Reliability Analysis (HRA) Fire PRA notebook (Ni-HRA-FO001) provides feasibility evaluations for the recovery actions modeled in the Fire PRA that are used to resolve VFDRs identified in the NSCA. The feasibility evaluations are performed separately for each of the 17 recovery actions identified for analysis considering the 11 criteria from FAQ 07-0030 (demonstrations, systems and indications, communications, emergency lighting, tools, procedures, staffing, actions in the fire area, time, training, and drills).* Recovery actions not modeled in the Fire PRA and whose additional risk was found to be insignificant based on a qualitative evaluation.

The feasibilities of these recovery actions are evaluated in EIR 51-9156521.

The recovery action feasibilities were evaluated using the 11 criteria given in FAQ 07-0030, which include, among others, staffing and timing requirements.

All recovery actions credited in the NFPA 805 transition were found to meet the feasibility criteria of FAQ 07-0030. Operator impacts and staffing considerations for the long-term safe and stable actions have been included in the updated 8 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)analysis.

Actions required to achieve and initially maintain hot shutdown conditions can be performed by the minimum shift complement of reactor operators, senior reactor operators, and non-licensed plant operators.

As discussed in the response to Part a of the RAI, EDG and DFP refueling activities do not occur for several days, are proceduralized, and can be implemented by emergency response organization personnel.

Operator actions modeled in the Fire PRA (including those that are not recovery actions) were evaluated for their feasibility and reliability as part of the development of their human error probability (HEP). This evaluation was documented in N1-HRA-F001 and relied on guidance from NUREG/CR-6850/EPRI TR-1011989, NUREG-1792, NUREG-1852, and the ASME/ANS PRA standard.The recovery actions that were previously credited in the fire risk evaluations to reduce the risk contribution from cold shutdown VFDRs are removed. These recovery actions are no longer required to demonstrate that safe and stable conditions can be maintained.

Recovery actions for the SDC system that are included in the Fire PRA as a means of reducing risk are retained.Part e The recovery actions involving local repairs to valve and pump wiring for the SDC system have been removed from LAR Attachment G. These recovery actions are associated with VFDRs pertaining to cold shutdown activities and are no longer required to demonstrate that safe and stable conditions (redefined as hot shutdown) can be maintained.

The VFDRs associated with these repair actions have been removed from the NSCA.Part f The risk of failure of operator actions and equipment necessary to sustain safe and stable conditions is evaluated in the models developed for the Fire PRA, since safe and stable conditions have been redefined as hot shutdown and the Fire PRA covers hot shutdown conditions.

Changes to the Fire PRA to implement the newly established safe and stable basis for the At-Power analysis involve credit for the DFP and two new operator recovery actions. The recovery actions are associated with manual valve alignment and local operation of the DFP to refill the Emergency Condenser makeup tanks in support of long-term operation of the EC system to satisfy decay heat removal requirements for hot shutdown.The Fire PRA models are quantified to determine the fire-induced core damage frequency (CDF) and large early release frequency (LERF). These risk metrics are used in the fire risk evaluations (FREs), consistent with Regulatory Guides 1.205 and 1.174.The Fire PRA also supports the FREs by ensuring that the risk inherent to each fire area is properly captured and that the set of recovery actions credited for the NFPA 805 transition is appropriately characterized, including the evaluation of their additional risk. LAR Attachment G is therefore amended to address the removal of SDC recovery actions and the addition of the DFP and associated valve-related recovery actions.9 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Updates to the LAR Attachment W Fire PRA insights and results reflect the Fire PRA re-quantification values generated for the revised safe and stable basis.Part g Multiple Spurious Operations (MSOs) associated with the SDC system are not addressed in the NSCA per the revised NMP1 basis for safe and stable operation.

As a result, VFDRs originally associated with the SDC system have been removed from the NSCA analysis.The NPO analysis (documented in EIR 51-9137629 and 51-9171174) identifies equipment that must remain functional to satisfy a particular Key Safety Function (KSF) success path. These KSF success paths were developed in accordance with the guidance in FAQ-07-0040, wherein higher risk evolutions (HRE) drive the selection of KSF success paths (including decay heat removal) based on time to boil. The NPO analysis provides recommendations to best manage fire risk for "pinch points" (areas of the plant where complete loss of a KSF may occur due to fire).A number of MSO scenarios associated with the SDC system are identified in the NMP1 Expert Panel MSO Report, "Technical Report on Identification

& Classification of the NMP-l MSO Scenarios using an Expert Panel -Review of New Generic Scenarios," dated May 2012. These scenarios are all addressed within the context of the SDC KSFs (1DHR-RX-SDC), as documented in the NMP1 NPO KSF Equipment List (EIR 51-9137629).

There is one KSF identified for each train of the SDC system. Accordingly, pinch points associated with the availability of these KSFs are identified in the NMP1 NPO Component Pinch Point Analysis (EIR 51-9171174).

Recommendations to best manage fire risk for each scenario pinch point are also described in EIR 51-9171174 (Appendix B) and are summarized in the response to Safe Shutdown / Circuit Analysis RAI 03 (see Table SSD/CA RAI 03-2).The response to Safe Shutdown / Circuit Analysis RAI 03 addresses specific aspects of the NMP1 NPO analysis, including the treatment of MSOs.10 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-1: Cold Shutdown VFDRs Eliminated from Table B-3 Fire Area Fire Area VFDR ID VFDR Title Details Comments Description 04 Foam Room, el. VFDR-04-005 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of 261 instrument air. BV-70-53 may fail closed on loss of the Shutdown Cooling System. The Shutdown instrument air. Valve is required open for cold shutdown to Cooling System is only required to support support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated required to open valve BV-70-53. (PC031) from the NSCA and will be addressed in the NPO analysis.04 Foam Room, el. VFDR-04-006 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of 261 instrument air. FCV-38-09 may fail closed on loss of the Shutdown Cooling System. The Shutdown instrument air. Valve is required open for cold shutdown to Cooling System is only required to support support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated required to open valve FCV-38-09. (PC031, OMC0O1) from the NSCA and will be addressed in the NPO analysis.04 Foam Room, el. VFDR-04-007 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of 261 instrument air. FCV-38-10 may fail closed on loss of the Shutdown Cooling System. The Shutdown instrument air. Valve is required open for cold shutdown to Cooling System is only required to support support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated required to open valve FCV-38-10. (PC031, OMC001) from the NSCA and will be addressed in the NPO analysis.04 Foam Room, el. VFDR-04-008 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of 261 instrument air. FCV-38-11 may fail closed on loss of the Shutdown Cooling System. The Shutdown instrument air. Valve is required open for cold shutdown to Cooling System is only required to support support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated required to open valve FCV-38-11. (PC031,OMC001) from the NSCA and will be addressed in the NPO analysis.05 Turbine VFDR-05-020 Loss of Instrument Air A deterministic assumption assumes potential loss of This VFDR is associated only with operation of Building, el. 240 instrument air. RBCLC to SDC Flow control valve BV-70-53 the Shutdown Cooling System. The Shutdown to 369 may fail closed on loss of instrument air. Valve is required Cooling System is only required to support open for cold shutdown to support decay heat removal. A CSD. On this basis, the VFDR is eliminated Recovery Action may be required to open valve BV-70-53 from the NSCA and will be addressed in the (PC031) NPO analysis.05 Turbine VFDR-05-021 Loss of Instrument Air A deterministic assumption assumes for potential loss of This VFDR is associated only with operation of Building, el. 240 instrument air. SDC heat exchanger outlet flow control the Shutdown Cooling System. The Shutdown to 369 valve FCV-38-10 may fail closed on loss of instrument air. Cooling System is only required to support Valve is required open for cold shutdown (PC031) CSD. On this basis, the VFDR is eliminated from the NSCA and will be addressed in the NPO analysis.05 Turbine VFDR-0S-025 Spurious Operation of A separation concern exists for postulated fire in this area This VFDR is associated only with operation of Building, el. 240 Shutdown Cooling for the loss of PB-167 due to an uncoordinated associated the Shutdown Cooling System. The Shutdown to 369 System Valves IV-38-01 emergency lighting circuit. Non SSD Emergency Lighting 11 Cooling System is only required to support and IV-38-13 is supplied by cable 167-101 from 600V Power Board 167, CSD. On this basis, the VFOR is eliminated Breaker H01. The emergency lighting power cable 167-101 from the NSCA and will be addressed in the supply breaker H01 does not coordinate with PB-167 supply NPO analysis.breaker. Fire damage to cable 167-101 could cause the loss of PB-167 preventing operation of Reactor Shutdown Cooling Isolation Valves, IV-38-01 and IV-38-13 as directed in repair procedure N1-DRP-005. ( OP019)05 Turbine VFDR-05-044 Unavailability of A postulated fire in this area may damage cable 171-41 This VFDR is associated only with operation of Building, el. 240 Shutdown Cooling adversely affecting credited SDC valve BV-38-04.

SDC is the Shutdown Cooling System. The Shutdown to 369 Valve BV-38-04 required to support the decay heat removal function.

Local- Cooling System is only required to support Manual operation of BV-38-04 per N1-SOP-21.1 may be CSD. On this basis, the VFDR is eliminated required. (PC031) from the NSCA and will be addressed in the NPO analysis.05 Turbine VFDR-05-045 Unavailability of A postulated fire in this area may damage cable 12DV-10, This VFDR is associated only with operation of Building, el. 240 Shutdown Cooling 12DV-11, 12DV-29, 12DV-9 or 167-11 adversely affecting the Shutdown Cooling System. The Shutdown to 369 Valve IV-38-02 credited SDC valve IV-38-02.

SDC is required to support the Cooling System is only required to support decay heat removal function.

Local-Manual operation of IV- CSD. On this basis, the VFDR is eliminated 38-02 per N1-SOP-21.1 may be required. (PC031) from the NSCA and will be addressed in the NPO analysis.11 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-1: Cold Shutdown VFDRs Eliminated from Table B-3 Fire Area Fire Area VFDR ID VFDR Title Details Comments Description 06 Turbine Building VFDR-06-013 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of North, el. 250 instrument air. Flow control valve BV-70-53 may fail closed the Shutdown Cooling System. The Shutdown on loss of instrument air. Valve is required open for cold Cooling System is only required to support shutdown to support decay heat removal. A Recovery CSD. On this basis, the VFDR is eliminated Action may be required to open valve BV-70-53. (PC031) from the NSCA and will be addressed in the NPO analysis.06 Turbine Building VFDR-06-014 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of North, el. 250 instrument air. SDC heat exchanger outlet flow control the Shutdown Cooling System. The Shutdown valve FCV-38-11 may fail closed on loss of instrument air. Cooling System is only required to support Valve is required open for cold shutdown to support decay CSD. On this basis, the VFDR is eliminated heat removal. A Recovery Action may be required to open from the NSCA and will be addressed in the valve FCV-38-11. (OMC001,PCO31)

NPO analysis.06 Turbine Building VFDR-06-015 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of North, el. 250 instrument air. SDC heat exchanger outlet flow control the Shutdown Cooling System. The Shutdown valve FCV-38-09 may fail closed on loss of instrument air. Cooling System is only required to support Valve is required open for cold shutdown to support decay CSD. On this basis, the VFDR is eliminated heat removal. A Recovery Action may be required to open from the NSCA and will be addressed in the valve FCV-38-09. (OMC001,PC031)

NPO analysis.06 Turbine Building VFDR-06-018 Unavailability of A postulated fire in this area may result in the loss of Train This VFDR is associated only with operation of North, el. 250 Shutdown Cooling 12 power adversely affecting credited SDC valve IV-38-02.

the Shutdown Cooling System. The Shutdown Valve IV-38-02 SDC is required to support the decay heat removal function.

Cooling System is only required to support Local-Manual operation of IV-38-02 per N1-SOP-21.1 may CSD. On this basis, the VFDR is eliminated be required. (PC031) from the NSCA and will be addressed in the NPO analysis.07 Turbine Building VFDR-07-003 Unavailability of A separation concern exists for a postulated fire in the area This VFDR is associated only with operation of South & West, Shutdown Cooling for shutdown cooling. Credited pump, PMP-38-152, is the Shutdown Cooling System. The Shutdown el. 250 Pump PMP-38-152 required to operate to support decay heat removal. Fire Cooling System is only required to support damage to cable 17-62 can prevent remote start of the CSD. On this basis, the VFDR is eliminated credited SDC pump due to loss of RCS Temperature Switch from the NSCA and will be addressed in the permissive.

Local breaker operation is required. (OPOIOA, NPO analysis.OMC001)07 Turbine Building VFDR-07-009 Loss of Instrument Air A deterministic assumption assumes a potential loss of This VFDR is associated only with operation of South & West, instrument air for a fire in any area of the plant. BV-70-53 the Shutdown Cooling System. The Shutdown el. 250 may fail closed on loss of instrument air. Valve is required Cooling System is only required to support open for cold shutdown to support decay heat removal. A CSD. On this basis, the VFDR is eliminated Recovery Action may be required to open valve BV-70-53 from the NSCA and will be addressed in the locally to supply RBCLC water to the SDC heat exchanger.

NPO analysis.(PCO31)07 Turbine Building VFDR-07-010 Loss of Instrument Air A deterministic assumption assumes a potential loss of This VFDR is associated only with operation of South & West, instrument air for a fire in any area of the plant FCV-38-10 the Shutdown Cooling System. The Shutdown el. 250 may fail closed on loss of instrument air. Valve is required Cooling System is only required to support open for cold shutdown to support decay heat removal. A CSD. On this basis, the VFDR is eliminated Recovery Action may be required to open valve FCV-38-10 from the NSCA and will be addressed in the locally to control SDC cooldown. (PC031) NPO analysis.09 Turbine Building VFDR-09-013 Loss of Instrument Air A deterministic assumption assumes a potential loss of This VFDR is associated only with operation of East, el. 250 instrument air for a fire in any fire area of the plant. Flow the Shutdown Cooling System. The Shutdown control valve BV-70-53 may fail closed on loss of instrument Cooling System is only required to support air. Valve is required open for cold shutdown to support CSD. On this basis, the VFDR is eliminated decay heat removal. A recovery action may be required to from the NSCA and will be addressed in the open valve BV-70-53 locally to supply RBCLC water to SDC NPO analysis.heat exchanger. (PC031)09 Turbine Building VFDR-09-014 Loss of Instrument Air A deterministic assumption assumes a potential loss of This VFDR is associated only with operation of East, el. 250 instrument air for a fire in any fire area of the plant. SDC the Shutdown Cooling System. The Shutdown heat exchanger outlet flow control valve FCV-38-11 may fail Cooling System is only required to support closed on loss of instrument air. Valve is required open for CSD. On this basis, the VFDR is eliminated cold shutdown to support decay heat removal. A Recovery from the NSCA and will be addressed in the Action may be required to open valve FCV-38-11 (PC031, NPO analysis.OMCO01)12 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-1: Cold Shutdown VFDRs Eliminated from Table B-3 Fire Area Fire Area VFDR ID VFDR Title Details Comments Description 09 Turbine Building VFDR-09-015 Loss of Instrument Air A deterministic assumption assumes a potential loss of This VFDR is associated only with operation of East, el. 250 instrument air for a fire in any fire area of the plant. SDC the Shutdown Cooling System. The Shutdown heat exchanger outlet flow control valve FCV-38-09 may fail Cooling System is only required to support closed on loss of instrument air. Valve is required open for CSD. On this basis, the VFDR is eliminated cold shutdown to support decay heat removal. A Recovery from the NSCA and will be addressed in the Action may be required to open valve FCV-38-09 (PC031, NPO analysis.OMC001)09 Turbine Building VFDR-09-021 Unavailability of A postulated fire in this area may result in the loss of Train This VFDR is associated only with operation of East, el. 250 Shutdown Cooling 12 power adversely affecting credited SDC valve IV-38-02.

the Shutdown Cooling System. The Shutdown Valve IV-38-02 SOC is required to support the decay heat removal function.

Cooling System is only required to support Local-Manual operation of IV-38-02 per N1-SOP-21.1 may CSD. On this basis, the VFDR is eliminated be required. (PC031) from the NSCA and will be addressed in the NPO analysis.10 Cable Spreading VFDR-10-011 Failure of Shutdown A separation concern exists for a postulated fire in this area This VFDR is associated only with operation of Room, el. 250-0 Cooling System Valve for the Shutdown Cooling system. SDC valve IV-38-01 is the Shutdown Cooling System. The Shutdown IV-38-01 To Open required open for CSD to support decay heat removal. Fire Cooling System is only required to support damage (ground) to cable 12DV-29 prevents SDC IV-38-01 CSD. On this basis, the VFDR is eliminated from opening. When power is restored to the valve and the from the NSCA and will be addressed in the control switch operated, the control circuit fuse will blow NPO analysis.due to a dead short across the CPT. The EC's will attempt to initiate automatically on either high reactor pressure or low-low reactor level. However, both paths of EC's may be adversely impacted by the following:

Potential inventory losses via the Main Steam Lines and EC vent and drain lines discussed above(OP039, OP048)10 Cable Spreading VFDR-10-019 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Room, el. 250-0 instrument air. BV-70-53 may fail closed on loss of the Shutdown Cooling System. The Shutdown instrument air. Valve is required open for cold shutdown to Cooling System is only required to support support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated required to open valve BV-70-53 locally to supply RBCLC from the NSCA and will be addressed in the water to SDC heat exchangers. (PC031) NPO analysis.10 Cable Spreading VFDR-10-020 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Room, el. 2S0-0 instrument air. FCV-38-09 may fail closed on loss of the Shutdown Cooling System. The Shutdown instrument air. Valve is required open for cold shutdown to Cooling System is only required to support support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated required to open valve FCV-38-09 locally to control SDC from the NSCA and will be addressed in the cooldown. (PC031) NPO analysis.10 Cable Spreading VFDR-10-026 Unavailability of A postulated fire in this area may damage cable 16-34 This VFDR is associated only with operation of Room, el. 250-0 Shutdown Cooling adversely impacting credited SDC pump PMP-38-149.

PMP- the Shutdown Cooling System. The Shutdown Pump PMP-38-149 38-149 is required to support decay heat removal. SDC Cooling System is only required to support PMP-38-149 is repaired and operated locally at PB 16 per CSD. On this basis, the VFDR is eliminated N1-DRP-GEN-003, Attachment

6. (PC037) from the NSCA and will be addressed in the NPO analysis.10 Cable Spreading VFDR-10-027 Unavailability of A postulated fire in this area may damage cable 12DV-11 or This VFDR is associated only with operation of Room, el. 250-0 Shutdown Cooling 12DV-29 adversely affecting credited SDC valve IV-38-02.

the Shutdown Cooling System. The Shutdown Valve IV-38-02 SDC is required to support the decay heat removal function.

Cooling System is only required to support Local-Manual operation of IV-38-02 per N1-SOP-21.1 may CSD. On this basis, the VFDR is eliminated be required. (PC031) from the NSCA and will be addressed in the NPO analysis.13 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-1: Cold Shutdown VFDRs Eliminated from Table B-3 Fire Area Fire Area VFDR ID VFDR Title Details Comments Description 11 Control VFDR-11-012 Unavailability of A separation concern exists for a postulated fire in this area This VFDR is associated only with operation of Complex, el. 261 Shutdown Cooling for shutdown cooling. Credited SDC pump, PMP-38-152, is the Shutdown Cooling System. The Shutdown and el. 277 Pump PMP-38-152 required to support decay heat removal for CSD. The Cooling System is only required to support credited SDC pump, PMP-38-152, may spuriously start and CSD. On this basis, the VFDR is eliminated run with no suction source. SDC pump PMP-38-152 may from the NSCA and will be addressed in the spuriously start due to an internal wire-to-wire short on NPO analysis.cable 17-23. SDC isolation valve IV-38-01 has power removed to prevent spurious opening for a fire in this fire area. SDC valves IV-38-02 and BV-38-04 are normally closed.Min-flow recirc valve FCV-38-131 may remain closed due to an internal wire-to-wire short on cable 1K-4. In the event the credited SDC pump is not available or other equipment operation causes a vessel overfill rendering EC's unavailable , various circuit failures in Train 11 and 12 CS valves could render the CS system unavailable for vessel injection and heat removal (OP044, OP026)11 Control VFDR-11-016 Loss of Instrument Air A deterministic assumption assumes a potential loss of This VFDR is associated only with operation of Complex, el. 261 instrument air. Block Valve, BV-70-S3 may fail closed on loss the Shutdown Cooling System. The Shutdown and el. 277 of instrument air. This Valve is required to be open for cold Cooling System is only required to support shutdown to support decay heat removal. A Recovery CSD. On this basis, the VFDR is eliminated Action may be required to open valve BV 53 (PC031) from the NSCA and will be addressed in the NPO analysis.11 Control VFDR-11-O24 Unavailability of Power A postulated fire in this area may cause the loss of PB 167 This VFDR is associated with a power supply Complex, el. 261 Board PB 167 due to damage to the following Train 11 component cables: recovery action to assure availability of the and el. 277 SDC system. On this basis, the VFDR is BKR-(16B/013B)R1043/603:

Cable 16-3 eliminated from the NSCA and will be BKR-(102/2-9)R1021/171:

Cables 101-6, 102-2, 111-10 addressed in the NPO analysis.BKR-(102/2-1)R1022/571:

Cables 101-6, 102-22, 102-28, 102-29, 102-30, 102-31, 102-43, 102-44, 102-45, 102-46, 102-49, 102-67, 11B-28, 1A-124, 1A-147, 1B-126, 1S-2386 BKR-(101/2B-1)R1012/151:

Cables 101-6, 11B-28, 1A-147, 1A-60 In support of shutdown from outside the control room, power is realigned from Train 12 to PB 167 per N1-DRP-GEN-004, Attachment

4. (PC044)11 Control VFDR-11-028 Unavailability of A postulated fire in this area may result in misoperation of This VFDR is associated only with operation of Complex, el. 261 Shutdown Cooling credited SDC isolation valve IV-38-01 due to fire damage to the Shutdown Cooling System. The Shutdown and el. 277 Valve IV-38-01 cable 167-11 or 167-12. To support meeting the decay heat Cooling System is only required to support removal function, SDC valve IV-38-01 is repaired and CSO. On this basis, the VFDR is eliminated operated from PB 167 per N1-DRP-GEN-004, Attachment from the NSCA and will be addressed in the 15. (PC049) NPO analysis.11 Control VFDR-11-029 Unavailability of A postulated fire in this area may result in misoperation of This VFDR is associated only with operation of Complex, el. 261 Shutdown Cooling credited SOC isolation valve IV-38-13 due to fire damage to the Shutdown Cooling System. The Shutdown and el. 277 Valve IV-38-13 cable 167-15 or 167-16. To support meeting the decay heat Cooling System is only required to support removal function, SDC valve IV-38-13 is repaired and CSD. On this basis, the VFDR is eliminated operated from PB 167 per N1-ORP-GEN-004, Attachment from the NSCA and will be addressed in the 16. (PCO5O) NPO analysis.11 Control VFDR-11-035 Manual Operation of In support of shutdown from outside the control room and This VFDR is associated only with operation of Complex, el. 261 Shutdown Cooling the decay heat removal performance function, SDC valves the Shutdown Cooling System. The Shutdown and el. 277 Valves BV-38-04, FCV- BV-38-04, FCV-38-10 and IV-38-02 are operated locally per Cooling System is only required to support 38-10, and IV-38-02 N1-DRP-GEN-004, Attachment 12, Attachment 13 or CSD. On this basis, the VFDR is eliminated Attachment
14. (PCO48) from the NSCA and will be addressed in the NPO analysis.12 Administration VFDR-12-005 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Building, el. instrument air. BV-70-53 may fail closed on loss of the Shutdown Cooling System. The Shutdown 250.0 instrument air. Valve is required open for cold shutdown to Cooling System is only required to support support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated required to open valve BV-70-S3. (PC031) from the NSCA and will be addressed in the NPO analysis.14 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAT 01-1: Cold Shutdown VFDRs Eliminated from Table B-3 Fire Area Fire Area VFDR ID VFDR Title Details Comments Description 12 Administration VFDR-12-006 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Building, el. instrument air. FCV-38-09 may fail closed on loss of the Shutdown Cooling System. The Shutdown 250.0 instrument air. Valve is required open for cold shutdown to Cooling System is only required to support support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated required to open valve FCV-38-09. (PC031) from the NSCA and will be addressed in the NPO analysis.12 Administration VFDR-12-007 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Building, el. instrument air. FCV-38-10 may fail closed on loss of the Shutdown Cooling System. The Shutdown 250.0 instrument air. Valve is required open for cold shutdown to Cooling System is only required to support support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated required to open valve FCV-38-10. (PC031) from the NSCA and will be addressed in the NPO analysis.12 Administration VFDR-12-008 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Building, el. instrument air. FCV-38-11 may fail closed on loss of the Shutdown Cooling System. The Shutdown 250.0 instrument air. Valve is required open for cold shutdown to Cooling System is only required to support support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated required to open valve FCV-38-11. (OMC001) from the NSCA and will be addressed in the NPO analysis.13 Screenhouse VFDR-13-005 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of instrument air for a fire in any fire area of the plant. FCV the Shutdown Cooling System. The Shutdown 09 may fail closed on loss of instrument air. Valve is Cooling System is only required to support required open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminated removal. A Recovery Action may be required to open valve from the NSCA and will be addressed in the FCV-38-09 (PC031, OMC001) NPO analysis.13 Screenhouse VFDR-13-006 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of instrument air for a fire in any fire area ofthe plant. FCV the Shutdown Cooling System. The Shutdown 10 may fail closed on loss of instrument air. Valve is Cooling System is only required to support required open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminated removal. A Recovery Action may be required to open valve from the NSCA and will be addressed in the FCV-38-10 (PC031, OMCO01) NPO analysis.13 Screenhouse VFDR-13-007 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of instrument air for a fire in any fire area ofthe plant. FCV the Shutdown Cooling System. The Shutdown 11 may fail closed on loss of instrument air. Valve is Cooling System is only required to support required open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminated removal. A Recovery Action may be required to open valve from the NSCA and will be addressed in the FCV-38-11. (PC031, OMCO01) NPO analysis.13 Screenhouse VFDR-13-008 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of instrument air for a fire in any fire area of the plant. BV the Shutdown Cooling System. The Shutdown 53 may fail closed on loss of instrument air. Valve is Cooling System is only required to support required open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminated removal. A Recovery Action may be required to open valve from the NSCA and will be addressed in the BV-70-53 (PC031) NPO analysis.14 Diesel Fire VFDR-14-005 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of pump Room, el. instrument air for a fire in any fire area of the plant. BV the Shutdown Cooling System. The Shutdown 261 53 may fail closed on loss of instrument air. Valve is Cooling System is only required to support required open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminated removal. A Recovery Action may be required to open valve from the NSCA and will be addressed in the BV-70-53. (PC031) NPO analysis.14 Diesel Fire VFDR-14-006 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of pump Room, el. instrument air for a fire in any fire area of the plant. FCV the Shutdown Cooling System. The Shutdown 261 09 may fail closed on loss of instrument air. Valve is Cooling System is only required to support required open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminated removal. A Recovery Action may be required to open valve from the NSCA and will be addressed in the FCV-38-09. (PC031,OMC001)

NPO analysis.14 Diesel Fire VFDR-14-007 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of pump Room, el. instrument air for a fire in any fire area of the plant. FCV the Shutdown Cooling System. The Shutdown 261 10 may fail closed on loss of instrument air. Valve is Cooling System is only required to support required open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminated removal. A Recovery Action may be required to open valve from the NSCA and will be addressed in the FCV-38-10. (PC031, OMH0O1) NPO analysis.15 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-1: Cold Shutdown VFDRs Eliminated from Table B-3 Fire Area Fire Area VFDR ID VFDR Title Details Comments Description 14 Diesel Fire VFDR-14-008 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of pump Room, el. instrument air for a fire in any fire area of the plant. FCV the Shutdown Cooling System. The Shutdown 261 11 may fail closed on loss of instrument air. Valve is Cooling System is only required to support required open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminated removal. A Recovery Action may be required to open valve from the NSCA and will be addressed in the FCV-38-11. (PC031,OMCO01)

NPO analysis.15 Radwaste and VFDR-15-005 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Waste Disposal instrument air. BV-70-53 may fail closed on loss of the Shutdown Cooling System. The Shutdown Buildings, el. instrument air. Valve is required open for cold shutdown to Cooling System is only required to support 252 to 292 support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated required to open valve BV-70-53. (PC031) from the NSCA and will be addressed in the NPO analysis.15 Radwaste and VFDR-1S-006 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Waste Disposal instrument air. FCV-38-09 may fail closed on loss of the Shutdown Cooling System. The Shutdown Buildings, el. instrument air. Valve is required open for cold shutdown to Cooling System is only required to support 252 to 292 support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated required to open valve FCV-38-09.(PC031, OMC001) from the NSCA and will be addressed in the NPO analysis.is Radwaste and VFDR-15-007 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Waste Disposal instrument air. FCV-38-10 may fail closed on loss of the Shutdown Cooling System. The Shutdown Buildings, el. instrument air. Valve is required open for cold shutdown to Cooling System is only required to support 252 to 292 support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated required to open valve FCV-38-10.(PCO31, OMC001) from the NSCA and will be addressed in the NPO analysis.15 Radwaste and VFDR-15-008 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Waste Disposal instrument air. FCV-38-11 may fail closed on loss of the Shutdown Cooling System. The Shutdown Buildings, el. instrument air. Valve is required open for cold shutdown to Cooling System is only required to support 252 to 292 support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated required to open valve FCV-38-11.(PCO31, OMCO01) from the NSCA and will be addressed in the NPO analysis.16A Battery Board VFDR-16A-00S Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Room 12, el. instrument air for a fire in any fire area of the plant. FCV the Shutdown Cooling System. The Shutdown 261 09 may fail closed on loss of instrument air. Valve is Cooling System is only required to support required open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminated removal. A Recovery Action may be required to open valve from the NSCA and will be addressed in the FCV-38-09. (PC031, OMC0O1) NPO analysis.16A Battery Board VFDR-16A-0O6 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Room 12, el. instrument air for a fire in any fire area of the plant. FCV the Shutdown Cooling System. The Shutdown 261 11 may fail closed on loss of instrument air. Valve is Cooling System is only required to support required open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminated removal. A Recovery Action may be required to open valve from the NSCA and will be addressed in the FCV-38-11. (PC031, OMC001) NPO analysis.16A Battery Board VFDR-16A-007 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Room 12, el. instrument air for a fire in any fire area of the plant. BV the Shutdown Cooling System. The Shutdown 261 53 may fail closed on loss of instrument air. Valve is Cooling System is only required to support required open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminated removal. A Recovery Action may be required to open valve from the NSCA and will be addressed in the BV-70-53. (PC031) NPO analysis.16A Battery Board VFDR-16A-008 Unavailability of Due to the unavailability of EDG 103 (Train 12, Path B), This VFDR is associated only with operation of Room 12, el. Shutdown Cooling power is not available to motor operated valve IV-38-02.

IV- the Shutdown Cooling System. The Shutdown 261 Valve IV-38-02 38-02 is a normally closed valve that fails as-is on loss of Cooling System is only required to support power. Valve is required open for cold shutdown to support CSD. On this basis, the VFDR is eliminated decay heat removal. A Recovery Action may be required to from the NSCA and will be addressed in the open valve IV-38-02. (PC031) NPO analysis.16B Battery Board VFDR-16B-005 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Room 11, el. instrument air for a fire in any fire area of the plant. FCV the Shutdown Cooling System. The Shutdown 261 10 may fail closed on loss of instrument air. Valve is Cooling System is only required to support required open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminated removal. A Recovery Action may be required to open valve from the NSCA and will be addressed in the FCV-38-10. (PC031) NPO analysis.16 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-1: Cold Shutdown VFDRs Eliminated from Table B-3 Fire Area Fire Area VFOR ID VFDR Title Details Comments Description 16B Battery Board VFDR-16B-006 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Room 11, el. instrument air for a fire in any fire area of the plant. BV the Shutdown Cooling System. The Shutdown 261 53 may fail closed on loss of instrument air. Valve is Cooling System is only required to support required open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminated removal. A Recovery Action may be required to open valve from the NSCA and will be addressed in the BV-70-53. (PC031) NPO analysis.17A Battery Room VFDR-17A-00S Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of 12, el. 277 to instrument air for a fire in any fire area of the plant. FCV the Shutdown Cooling System. The Shutdown 291 09 may fail closed on loss of instrument air. Valve is Cooling System is only required to support required open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminated removal. A Recovery Action may be required to open valve from the NSCA and will be addressed in the FCV-38-09.

NPO analysis.17A Battery Room VFDR-17A-006 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of 12, el. 277 to instrument air for a fire in any fire area of the plant. FCV the Shutdown Cooling System. The Shutdown 291 11 may fail closed on loss of instrument air. Valve is Cooling System is only required to support required open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminated removal. A Recovery Action may be required to open valve from the NSCA and will be addressed in the FCV-38-11.

NPO analysis.17A Battery Room VFDR-17A-007 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of 12, el. 277 to instrument air for a fire in any fire area of the plant. BV the Shutdown Cooling System. The Shutdown 291 53 may fail closed on loss of instrument air. Valve is Cooling System is only required to support required open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminated removal. A Recovery Action may be required to open valve from the NSCA and will be addressed in the BV-70-53.

NPO analysis.17A Battery Room VFDR-17A-O08 Unavailability of Due to the unavailability of EDG 103 (Train 12, Path B), This VFDR is associated only with operation of 12, el. 277 to Shutdown Cooling power is not available to motor operated valve IV-38-02.

IV- the Shutdown Cooling System. The Shutdown 291 Valve iV-38-02 38-02 is a normally closed valve that fails as-is on loss of Cooling System is only required to support power. Valve is required open for cold shutdown to support CSD. On this basis, the VFDR is eliminated decay heat removal. A Recovery Action may be required to from the NSCA and will be addressed in the open valve iV-38-02.

NPO analysis.178 Battery Room VFDR-17B-OOS Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of 11, el. 277 to instrument air for a fire in any fire area of the plant. FCV the Shutdown Cooling System. The Shutdown 291 10 may fail closed on loss of instrument air. Valve is Cooling System is only required to support required open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminated removal. A Recovery Action may be required to open valve from the NSCA and will be addressed in the FCV-38-10. (PC031) NPO analysis.17B Battery Room VFDR-17B-006 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of 11, el. 277 to instrument air for a fire in any fire area of the plant. BV the Shutdown Cooling System. The Shutdown 291 53 may fail closed on loss of instrument air. Valve is Cooling System is only required to support required open for cold shutdown to support decay heat CSD. On this basis, the VFDR is eliminated removal. A Recovery Action may be required to open valve from the NSCA and will be addressed in the BV-70-S3. (PC031) NPO analysis.18 Emergency VFDR-18-007 Flow Diversion from A spurious actuation concern exists impacting valves FCV- This VFDR is associated only with operation of Diesel Containment Spray 93-74 and FCV-93-73 by diverting flow due to wire-to-wire the CTS/CTRWS, which provides torus Generator 102 Raw Water System to shorts on the following cables. An internal wire-to-wire cooling. Torus cooling is required to support Missile Containment Spray short on cable 171-163 spuriously opens FCV-93-74 CSD, and is not necessary when the primary Enclosure, el. System diverting CTSRW flow to the CS system. (OP035) An internal decay heat removal method is achieved via 271 wire-to-wire short on cable 171-160 spuriously opens FCV- the EC's. On this basis, the VFDR is eliminated 93-73 diverting CTSRW flow to the CTS system. (OP034)19 Diesel VFDR-19-003 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Generator instrument air. BV-70-53 may fail closed on loss of the Shutdown Cooling System. The Shutdown Room 103, el. instrument air. Valve is required open for cold shutdown to Cooling System is only required to support 250 support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated required to open valve BV-70-53. (PC031) from the NSCA and will be addressed in the NPO analysis.19 Diesel VFDR-19-004 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Generator instrument air. FCV-38-11 may fail closed on loss of the Shutdown Cooling System. The Shutdown Room 103, el. instrument air. Valve is required open for cold shutdown to Cooling System is only required to support 250 support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated required to open valve FCV-38-11. (OMC001) from the NSCA and will be addressed in the NPO analysis.17 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-1: Cold Shutdown VFDRs Eliminated from Table B-3 Fire Area Fire Area VFDR ID VFDR Title Details Comments Description 19 Diesel VFDR-19-00S Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Generator instrument air. FCV-38-09 may fail closed on loss of the Shutdown Cooling System. The Shutdown Room 103, el. instrument air. Valve is required open for cold shutdown to Cooling System is only required to support 250 support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated required to open valve FCV-38-09. (PC031) from the NSCA and will be addressed in the NPO analysis.19 Diesel VFDR-19-O06 Unavailability of Due to the unavailability of EDG 103 (Train 12, Path B), This VFDR is associated only with operation of Generator Shutdown Cooling power is not available to motor operated valve IV-38-02.

IV- the Shutdown Cooling System. The Shutdown Room 103, el. Valve IV-38-02 38-02 is a normally closed valve that fails as-is on loss of Cooling System is only required to support 250 power. Valve is required open for cold shutdown to support CSD. On this basis, the VFDR is eliminated decay heat removal. A Recovery Action may be required to from the NSCA and will be addressed in the open valve IV-38-02. (PC031) NPO analysis.20 Diesel VFDR-20-001 Unavailability of Due to the unavailability of EDG 103 (Train 12, Path B), This VFDR is associated only with operation of Generator Shutdown Cooling power is not available to motor operated valve IV-38-02.

IV- the Shutdown Cooling System. The Shutdown Enclosed Valve IV-38-02 38-02 is a normally closed valve that fails as-is on loss of Cooling System is only required to support Cableway, el. power. Valve is required open for cold shutdown to support CSD. On this basis, the VFDR is eliminated 250 decay heat removal. A Recovery Action may be required to from the NSCA and will be addressed in the open valve IV-38-02. (PC031) NPO analysis.20 Diesel VFDR-20-004 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Generator instrument air. BV-70-53 may fail closed on loss of the Shutdown Cooling System. The Shutdown Enclosed instrument air. Valve is required open for cold shutdown to Cooling System is only required to support Cableway, el. support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated 250 required to open valve BV-70-53. (PC031) from the NSCA and will be addressed in the NPO analysis.20 Diesel VFDR-20-005 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Generator instrument air. FCV-38-11 may fail closed on loss of the Shutdown Cooling System. The Shutdown Enclosed instrument air. Valve is required open for cold shutdown to Cooling System is only required to support Cableway, el. support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated 250 required to open valve FCV-38-11. (OMC001) from the NSCA and will be addressed in the NPO analysis.20 Diesel VFDR-20-006 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Generator instrument air. FCV-38-09 may fail closed on loss of the Shutdown Cooling System. The Shutdown Enclosed instrument air. Valve is required open for cold shutdown to Cooling System is only required to support Cableway, el. support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated 250 required to open valve FCV-38-09. (PC031) from the NSCA and will be addressed in the NPO analysis.21 Below Power VFDR-21-001 Unavailability of Due to the unavailability of EDG 103 (Train 12, Path B), This VFDR is associated only with operation of Boards 102/103, Shutdown Cooling power is not available to motor operated valve IV-38-02.

IV- the Shutdown Cooling System. The Shutdown el. 250 Valve IV-38-02 38-02 is a normally closed valve that fails as-is on loss of Cooling System is only required to support power. Valve is required open for cold shutdown to support CSD. On this basis, the VFDR is eliminated decay heat removal. A Recovery Action may be required to from the NSCA and will be addressed in the open valve IV-38-02. (PC031) NPO analysis.21 Below Power VFDR-21-004 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Boards 102/103, instrument air. BV-70-53 may fail closed on loss of the Shutdown Cooling System. The Shutdown el. 2S0 instrument air. Valve is required open for cold shutdown to Cooling System is only required to support support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated required to open valve BV-70-53. (PC031) from the NSCA and will be addressed in the NPO analysis.21 Below Power VFDR-21-O05 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Boards 102/103, instrument air. FCV-38-11 may fail closed on loss of the Shutdown Cooling System. The Shutdown el. 250 instrument air. Valve is required open for cold shutdown to Cooling System is only required to support support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated required to open valve FCV-38-11. (OMC001, PC031) from the NSCA and will be addressed in the NPO analysis.21 Below Power VFDR-21-006 Loss of Instrument Air A deterministic assumption exists for potential loss of This VFDR is associated only with operation of Boards 102/103, instrument air. FCV-38-09 may fail closed on loss of the Shutdown Cooling System. The Shutdown el. 250 instrument air. Valve is required open for cold shutdown to Cooling System is only required to support support decay heat removal. A Recovery Action may be CSD. On this basis, the VFDR is eliminated required to open valve FCV-38-09.( OMC001,PC031) from the NSCA and will be addressed in the NPO analysis.18 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-1: Cold Shutdown VFDRs Eliminated from Table B-3 Fire Area Fire Area VFDR ID VFDR Title Details Comments Description 22 Emergency VFDR-22-006 Flow Diversion from A separation concern exists for a postulated fire in the area This VFDR is associated only with operation of Diesel Containment Spray for the Containment Spray Raw Water system. CTSRW is the Shutdown Cooling System. The Shutdown Generator 102 Raw Water System to required to support the decay heat removal function.

Both Cooling System is only required to support Foundation Containment Spray loops of the credited secondary decay heat removal CSD. On this basis, the VFDR is eliminated Room, el. 250 System function can be lost. An internal wire-to-wire short on cable from the NSCA and will be addressed in the and Diesel 171-163 spuriously opens FCV-93-74 diverting flow from NPO analysis.Generator pump PMP-93-03 to the CS system. An internal wire-to-wire Room, el. 261 short on cable 171-160 spuriously opens FCV-93-73 diverting flow from pump PMP-93-04 to the CTS system.Diversion of the CTSRW flow paths away from the CTSRW heat exchangers results in a loss of Decay Heat Removal.(OP034, OP035)19 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAT 01-2: New VFDRs Added to Table B-3 to Support Extended Hot Shutdown Operation Fire Area VFDR ID VFDR Title Details Comments The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 04 VFDR-04-009 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeu isneededfromtheFireProtection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and make m t he availabilityion 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 05 VFDR-05-047 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 06 VFDR-06-019 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 07 VFDR-07-014 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 09 VFDR-09-022 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 10 VFDR-10-029 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.20 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-2: New VFDRs Added to Table B-3 to Support Extended Hot Shutdown Operation Fire Area VFDR ID VFDR Title Details Comments The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 11 VFDR-11-037 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeu is needed from the Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and make m toe availabilityion 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 12 VFDR-12-009 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 13 VFDR-13-011 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 14 VFDR-144009 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1s VFDR4lS-O09 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 16A VFDR-16A-009 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is neededfromthe Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.21 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-2: New VFDRs Added to Table B-3 to Support Extended Hot Shutdown Operation Fire Area VFDR ID VFDR Title Details Comments The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of g hours 16B VFDR-16B-007 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 17A VFDRJ17A-009 Long-Term EC Make-up Tank water tanks isnecessaryto assure continued availability ofthe EC makeup is needed from the Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 17B VFDR-17B-007 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 18 VFDR-18-011 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability ofthe EC makeup is needed from the Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory mTakeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 19 VFDR-19-007 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 20 VFDR-20-008 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability ofthe EC makeup is neededfromthe Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a tiSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.22 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 01-2: New VFDRs Added to Table B-3 to Support Extended Hot Shutdown Operation Fire Area VFDR ID VFDR Title Details Comments The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 21 VFDR-21-008 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 22 VFDR-22-008 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 23 VFDR-23-008 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.The Emergency Condensers (ECs) are credited for establishing long-term decay heat removal to maintain plant HSD conditions.

After a This VFDR is associated with inventory period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the initiation of plant shutdown, inventory makeup to the Emergency Condenser makeup from the Fire Protection Water System to the EC makeup Makeup Tanks. After a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 24 VFDR-24-007 Long-Term EC Make-up Tank water tanks is necessary to assure continued availability of the EC makeup is needed from the Fire Protection Water Supply System. Valves 100-68 (loop 111 and 112) and 100-69 (loop 121 and System to assure long-term availability of 122) are normally closed manual valves, and are both required open the EC system. On this basis, this VFDR is to periodically refill the EC condenser makeup tanks. Additionally, classified a HSD VFDR.availability of the Diesel Driven Fire Pump (DFP) from the Main Control Room is not assured.23 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Safe Shutdown / Circuit Analysis RAI 03 LAR Section 4.3 and Attachment D describe the methods and results of the non-power operations (NPO)evaluation, including references to the applicable outage programs, procedures, and NPO analyses.Additional information is requested asfollows:

a. Provide "Appendix B: NMPJ NPO Pinch Point Assessment" in the NPO fire area reviews including a summary level identification of unavailable paths in each fire area and the resolution for each pinch point.b. During NPO modes, spurious actuation of valves can have a significant impact on the ability to maintain decay heat removal and inventory control. Provide a description of any actions being credited to minimize the impact of fire-induced spurious actuations on power operated valves (e.g., air operated valves (AOVs) and motor operated valves (MO Vs)) during NPO either as pre-fire conditioning or as required during the fire response recovery (e.g., pre-fire rack-out, locally pinning of valves, and isolation of air supplies).

For example, it appears to the NRC staff that the Technical Specifications (TS) allow the shutdown cooling isolation valves 38-01 and 38-13 to be inoperable in the open position for greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> under certain specific conditions.

During higher risk evolutions such as a short time to boil, preventing the spurious closure of any of these valves would be advantageous.

Provide justification for not invoking the TS allowed flexibility for maintaining these valves open during higher risk evolutions (HREs).c. Identify locations where key safety functions (KSFs) are achieved via RAs or for which instrumentation not already included in the at-power analysis is needed to support RAs required to maintain safe and stable conditions.

Identify those RAs and instrumentation relied upon in NPO and describe how RA feasibility is evaluated.

Include in the description whether these variables have been or will be factored into operator procedures supporting these actions.For instance, during outage conditions when there is a short time to boil, describe the operator response to a spurious closure of one of the shutdown cooling system motor operated isolation valves 38-01 or 38-13. Describe how any RAs are feasible (e.g., can be reliably accomplished in the available time frame).Response to Safe Shutdown / Circuit Analysis RAI 03 General The following is background information and other details of the non-power operations (NPO) analysis that form the baseline for the specific responses to Parts a through c of this RAI.NMPNS has elected to modify its NFPA 805 transition analysis for NMP1 to revise the approach for demonstrating the ability to reach and maintain safe and stable conditions, as specified by NFPA 805. The original Nuclear Safety Capability Assessment (NSCA) established as its basis for demonstrating safe and stable conditions the requirement to maintain Keff < 0.99 with a reactor coolant temperature at or below the requirements for hot shutdown and then subsequently cool down and maintain the plant in a cold 24 of 47 ENCLOSURE1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)shutdown condition.

Consistent with NFPA 805 and supplemental guidance, NMPNS is revising its basis for the NMP1 NSCA to include only the requirement to establish hot shutdown conditions, including long-term hot shutdown capability.

The impact of this change is primarily limited to the NSCA, which is addressed in the response to Safe Shutdown / Circuit Analysis RAI 01. The change to safe and stable conditions does not impact the NPO Plant Operating States (POSs), Key Safety Functions (KSFs), or pinch point analysis.

Hence, the response to this RAI does not depend on results or conclusions described in the response to Safe Shutdown / Circuit Analysis RAI 01.As discussed in the response to Safe Shutdown / Circuit Analysis RAI 01, demonstration of the nuclear safety performance criteria for safe and stable conditions is performed in two analyses based on the plant operating modes, as defined in the NMP1 TS. These analyses are defined as follows:* At-Power analysis for potential fires while in either: (i) the Power Operating Condition (Reactor mode switch is in "Startup" or "Run" position and the reactor is critical or criticality is possible due to control rod withdrawal), or (ii) the Shutdown Condition

-Hot operating condition (Reactor mode switch is in "Shutdown" position and reactor coolant temperature is greater than 212'F), with the Shutdown Cooling (SDC) system not aligned for decay heat removal. (Refer to the response to Safe Shutdown / Circuit Analysis RAI 01 for further discussion of this analysis and its results.)* Non-Power analysis for potential fires while in Shutdown Condition

-Hot operating condition and lower operating conditions.

A copy of TS Section 1.1 containing the definitions of the NMP 1 reactor operating conditions is provided as Figure SSD/CA RAI 01-1 in the response to Safe Shutdown / Circuit Analysis RAI 01. Table SSD/CA RAI 03-1 below provides a correlation between the three POSs identified in FAQ 07-0040 and plant operating modes defined in the NMPI TS. Note that the reference to "RHR" in the FAQ 07-0040 descriptions of POS is analogous to the SDC system at NMP 1.25 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 03-1: POS to TS Operating Condition Correlation POS Number and Description (from FAQ 07-0040)NMP1 TS Operating Condition and Description" This POS starts when the RHR -Reactor mode switch is in system is placed into service. "Shutdown" or "Refueling"" The vessel head is on and the position RCS is closed such that an m No core alterations leading P05 1 extended loss of the decay heat Shutdown Condition

-Hot to an addition of reactivity removal (DHR) function without Shutdown Condition

-Cold are being performed operator intervention could result m Reactor coolant temperature in a RCS re-pressurization above is greater than 212'F (Hot)the shutoff head for the RHR or equal to or less than pumps. 212°F (Cold)This POS represents the shutdown condition when: (1) The vessel head is removed and reactor pressure vessel water level is less than the POS 2 minimum level required for Major Maintenance No fuel is in the Reactor movement of irradiated fuel Condition assemblies within the reactor pressure vessel as defined by Technical Specifications, OR (2) A sufficient RCS vent path exists for decay heat removal.m This POS represents the -Reactor mode switch is in shutdown condition when the "Refueling" position reactor pressure vessel water m Fuel may be loaded or level is equal or greater than the unloaded POS 3 minimum level required for Refueling Condition m Reactor coolant temperature movement of irradiated fuel is less than 212'F assemblies within the reactor -No more than one operable pressure vessel as define by control rod is withdrawn Technical Specifications m This POS occurs during Mode 5 As described in LAR Attachment D, procedure NIP-OUT-01, "Shutdown Safety," defines higher risk evolutions (HREs) and establishes KSFs and defense-in-depth (DID) strategies to protect the KSFs. HREs are defined as: "Outage activities, plant configurations, or conditions during shutdown where the plant is more susceptible to an event causing the loss of a key safety function or the number of key safety systems drops below the shutdown safety criteria." NIP-OUT-01 ensures that HREs are identified and communicated to plant personnel with applicable precautions and / or contingency plans clearly identified; e.g., on the Outage Schedule Shutdown Safety Review (SSR) reports. KSFs considered for HREs, as required by NIP-OUT-01, include: 26 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)1. Decay Heat Removal Capability.

Assessments for maintenance activities affecting decay heat removal capability should consider that the ability of systems and components to remove decay heat is dependent on a variety of factors, including the plant configuration, availability of other key safety systems and components, and the ability of operators to diagnose and respond properly to an event.For example, assessment of maintenance activities that impact the decay heat removal key safety function should consider:* Initial magnitude of decay heat.* Time to boil.* Time to core uncover.* Initial RCS water inventory condition (for example, filled, reduced, reactor cavity flooded, etc.).* RCS configurations (for example, reactor vessel open/closed, recirculation nozzle plugs installed or loop isolation valves closed, vent paths available, temporary covers installed, main steam line plugs installed, etc.).When any fuel is offloaded to the spent fuel pool during the refueling outage, the decay heat removal function will be at least partially shifted from the RCS to the spent fuel pool (SFP). When the core is completely offloaded with the SFP gates installed, the decay heat removal function in the RCS can be marked "Not Applicable." 2. Inventory Control. Assessments for maintenance activities should address the potential for creating inventory loss flow paths in both the RCS and the SFP. For example: Maintenance activities associated with the main steam lines (for example, safety or relief valve removal, automatic depressurization system testing, main steam isolation valve maintenance, and so forth) can create a drain down path for the reactor cavity and fuel pool. This potential is significantly mitigated through the use of main steam plugs.When the core is completely offloaded with the SFP gates installed, the reactor Inventory Control function can be marked "Not Applicable." 3. Power Availability.

Assessments should consider the impact of maintenance activities on availability of electrical power. Electrical power is required during shutdown conditions to maintain cooling to the reactor core and the SFP, to transfer decay heat to the heat sink, to achieve containment closure when needed, and to support other important functions.

  • Assessments for maintenance activities involving AC power sources and distribution systems should address providing defense in depth that is commensurate with the plant operating mode or configuration.
  • Assessments for maintenance activities involving the switchyard and transformer yard should consider the impact on offsite power availability.

27 of 47 ENCLOSURE1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)AC and DC instrumentation and control power is required to support systems that provide key safety functions during shutdown.

As such, maintenance activities affecting power sources, inverters, or distribution systems should consider their functionality as an important element in providing appropriate defense in depth.4. Reactivity Control. The main aspect of this key safety function involves maintaining adequate shutdown margin in the reactor core and the SFP. During periods of cold weather, RCS temperatures can also decrease below the minimum value assumed in the shutdown margin calculation.

When in power operation or startup conditions, availability of the Liquid Poison system must be considered.

KSFs identified in NIP-OUT-01 associated with POSs specifically excluded from consideration by FAQ 07-0040 are not discussed in the response to this RAI.Updates to Section 4.3 and Attachment F of the LAR Transition Report that are associated with the response to this RAI are provided in Enclosures 3 and 4.Part a Table SSD/CA RAI 03-2 below, from EIR 51-9171174, Appendix B -NMP1 NPO Pinch Point Assessment, provides summary level identification of KSF losses and pinch points on a fire zone basis. The table identifies each KSF associated with a pinch point and the recommendations for addressing the pinch points.As described in Section 4.3 and Attachment D of the LAR, the following KSFs are evaluated in each fire zone: " Decay Heat Removal (DHR) for both the Reactor Vessel (RX) and the Spent Fuel Pool (SFP).* Inventory Control (INV) for both the Reactor Vessel and the Spent Fuel Pool.* Power (PWR) availability.

The Reactivity Control KSF is not included in the NPO analysis because it is administratively controlled in accordance with procedure NIP-OUT-01.

Referring to Table SSD/CA RAI 03-2, the KSFs are categorized with codes assigned to each KSF -Fire Zone pair. Three codes have been established to summarize the fire impacts: ' "1" (Impacted):

At least one of the KSF paths associated with a given KSF is affected; i.e., a component of a specific KSF path or any of the component's required cables within the fire zone are impacted, whereby that path can no longer be assured of being functional.

However, at least one other KSF path for the KSF remains available.

  • "L" (Lost): All available success paths for a given KSF are impacted.* "N" (None): No impacts to the KSF are identified.

28 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)"Pinch Points" are then identified (on a fire zone basis), based on the loss of a KSF. An "N" in the pinch point column of Table SSD/CA RAI 03-2 indicates that no KSFs are lost in this fire zone. A"Y" in this column indicates that one or more KSFs are potentially lost in the fire zone, and therefore a pinch point is considered to exist. Fire zones are then categorized as follows:* Category 1 fire zones are not pinch points as they were found to have at least one success path for each KSF. No recommendations for additional fire protection measures during HREs are made for these zones. Standard DID strategies, as specified by procedure NIP-OUT-01, "Shutdown Safety," are adequate to address risk." Category 2 fire zones are pinch points as every success path is potentially lost for at least one KSF. These KSF success paths can be preserved through fire protection/fire prevention actions, including the verification of functionality of available fire detection and suppression during HREs.FAQ 07-0040 provides a listing of standard fire risk management methods that have been found to be acceptable for managing fire risk during HREs. During periods of NPO that are not defined as HREs, the standard fire protection DID actions are considered sufficient to minimize fire risk. During HREs, recommendations from FAQ 07-0040 have been identified for additional measures to consider as part of a comprehensive program to reduce fire risk. Each Category 2 fire zone includes one or more recommendations from the list provided in Table SSD/CA RAI 03-3 to minimize fire risk to the KSFs, as described in Table SSD/CA RAI 03-2. Note that Recommendations 2B, 4, 6, and 7 from FAQ 07-0040 are not used.29 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING TILE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 03-2: Summary Level Identification of KSF Losses and Pinch Points (from EIR 51-9171174, Appendix B -NMP1 NPO Pinch Point Assessment)

Fire KSFs Lost or Impacted Pinch Recommendations Fire Zone Area Fire Zone Description DHR INV Point Category (from Table SSD/CA RAI 03-3) Suppression Detection AraPWR ?RX SFP RX SFP Al NA ADMINISTRATION BUILDING I I I I I N 1 Not a pinch point. No action needed.EL 250-0 A2 NA ADMINISTRATION BUILDING A2 NA EL 248-0 I I I I I N 1 Not a pinch point. No action needed.ABIA 12 RECORDS STORAGE AREA EL N N N N N N 1 Not a pinch point. No action needed.250-0 ABIB 12 SAS EQUIPMENT AREA EL N N N N N N 1 Not a pinch point. No action needed.252-0 AB1C 12 CPU EQUIPMENT AREA EL I I I I I N 1 Not a pinch point. No action needed.252-0 ABID 12 GENERAL AREA EL 250-0 I I I I I N 1 Not a pinch point. No action needed.ABlIE 12 LOCKER AREA, LUNCH ROOM, N N N N N N 1 Not a pinch point. No action needed.OFFICES EL 261-0 AB1F 4 FOAM ROOM EL 261-0 I I I I I N 1 Not a pinch point. No action needed.AB2A 12 ACCESS PASSAGEWAY EL N N N N N N 1 Not a pinch point. No action needed.248-0 TECHNICAL SUPPORT AREA AB2B 12 EL 248-0 N N N N N N 1 Not a pinch point. No action needed.AB2C 12 RADIATION RECORDS AREA N N N N N N 1 Not a pinch point. No action needed.EL 248-0 AB2D 12 WAREHOUSE AREA EL 248-0 N N N N N N 1 Not a pinch point. No action needed.AB3A 12 WAREHOUSE AREA EL 261-0 N N N N N N 1 Not a pinch point. No action needed.AB3B 12 OIL STORAGE ROOM EL 261-0 N N N N N N 1 Not a pinch point. No action needed.AB3C 12 STOREROOM TRUCK DOCK N N N N N N 1 Not a pinch point. No action needed.EL 261-0 ELECTRICAUMECHANICAL AB3D 12 SHOP AREA, OFFICE AREAS, N N N N N N 1 Not a pinch point. No action needed.LOCKER ROOMS EL 261-0 30 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)KSFs Lost or Impacted Pinch Recommendations Fire Zone Fre Fire Zone Description DHR INV Point Category (from Table SSDICA RAI 03-3) Suppression Detection AraPWR ?RX SFP RX SFP AB3E 12 TELEPHONE ROOM 1 EL 261-0 N N N N N N 1 Not a pinch point. No action needed.AB3F 12 TELEPHONE ROOM 2 EL 261-0 N N N N N N 1 Not a pinch point. No action needed.AB4A 12 GENERAL OFFICE AREA EL N N N N N N 1 Not a pinch point. No action needed.277-0 AB4B 12 FILE ROOM EL 277-0 N N N N N N 1 Not a pinch point. No action needed.AB4C 12 RECORDS PROCESSING N N N N N N 1 Not a pinch point. No action needed.AREA EL 277-0 1 AB4D 12 GENERAL OFFICE AREA EL N N N N N N 1 Not a pinch point. No action needed.277-0 PENTHOUSE VENTILATION AB5 12 ROOM EL 290-0 N N N N N N 1 Not a pinch point. No action needed.ROOMTERY BOA90 OO-10E B1A 16A BATTERY BOARD ROOM 12 EL L L L I Y 2 1A and/or 3B and/or 5 None Yes 261-0 B1B 16B BATTERY BOARD ROOM 11 EL L L I I Y 2 1A and/or 3B and/or 5 None Yes 261-0 B2A 17A BATTERY ROOM 12 EL 277-0 I L L L I Y 2 1A and/or 3B and/or 5 None Yes B2B 17B BATTERY ROOM 11 EL 277-0 I L L I I Y 2 1A and/or 3B and/or 5 None Yes C1 10 CABLE SPREADING ROOM EL L L L L L Y 2 1A and/or 2A and/or 3A Yes Yes 250-0 C2 11 AUXILIARY CONTROL ROOM, L L L L L Y 2 1A and/or 2A and/or 3A and/or 5 Yes Yes COMPUTER ROOM 261-0 C3 11 CONTROL ROOM EL 277-0 L L L L L Y 2 9 Yes Yes EDG 103 FOUNDATION ROOM D1A 19 EL 20 I L I L I Y 2 IA and/or 2A and/or 3A Yes Yes EL 250-0 D1B 22 EG12FUDTOROM I L L I I Y 2 1A and/or 2A and/or 3A Yes Yes EL 250-0 D1C 20 EDG 103 CABLE ROUTING I L L L L Y 2 1A and/or 2A and/or 3A Yes Yes AREA EL 250-0 DID 21 ROOM BELOW PB'S 102& 103 I L L L L Y 2 1A and/or 2A and/or 3A Yes Yes EL 250-0 D2A 19 EDG 103 ROOM EL 261-0 I L I L I Y 2 1A and/or 2A and/or 3A Yes Yes 31 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)FireZone Fire FiKSFs Lost or Impacted Pinch Recommendations F nea DHR INV Point Category (from Table SSD/CA RAI 03-3) Suppression Detection RX SFP RX SFP D2B 22 EDG 102 ROOM EL 261-0 L L L L I Y 2 1A and/or 2A and/or 3A Yes Yes D2C 23 POWER BOARD 102 ROOM EL L L L L I Y 2 1A and/or 2A and/or 3A Yes Yes 261-0 D2D 24 POWERBOARD103ROOMEL L L L L I Y 2 1A and/or 2A and/or 3A Yes Yes 261-0 EDG 102 CONTROL CABLE D3 18 MISSILE ENCLOSURE EL 271- L L L L I Y 2 1A and/or 5 None Yes 0 EXT EXT EXTERNAL TO PLANT L L L L L Y 2 1A and/or 2A and/or 3A and/or 10 Yes Yes F1 4 FOAM ROOM EL 261-0 N N N N N N 1 Not a pinch point. No action needed.FBZ-1 1 REACTOR BUILDING FIRE N N N N N N 1 Not a pinch point. No action needed.BREAK ZONE FBZ-2 2 REACTOR BUILDING FIRE N N N N N N 1 Not a pinch point. No action needed.BREAK ZONE FBZR237N-1 1 REACTOR BUILDING EL 237-0 I I I I N 1 Not a pinch point. No action needed.COL N-Q, ROW 8-9 FBZR237N-2 2 REACTOR BUILDING EL 237-0 1 1 1 1 N 1 Not a pinch point. No action needed.COL N-Q, ROW 8-9 FBZR261N-1 1 REACTOR BUILDING EL 261-0 L L I I I Y 2 1Aand/or3Band/or5 Yes None COL N-Q, ROW 8-9 FBZR261N-2 2 REACTOR BUILDING EL 261-0 L L I I I Y 2 1A and/or 3B and/or 5 Yes None COL N-Q, ROW 8-9 FBZR281N-1 1 REACTOR BUILDING EL 281-0 L L I I I Y 2 1A and/or 3B and/or 5 Yes None COL M-Q, ROW 6-7 FBZR281N-2 2 REACTOR BUILDING EL 281-0 L L I I I Y 2 1A and/or 3B and/or 5 Yes None COL M-0, ROW 6-7 FBZR281S-1 1 REACTOR BUILDING EL 281-0 1 1 1 1 N 1 Not a pinch point. No action needed.COL K-L, ROW 7-8 FBZR281 S-2 2 REACTOR BUILDING EL 281-0 1 1 1 1 N 1 Not a pinch point. No action needed.COL K-L, ROW 7-8 FBZR298N-1 1 REACTOR BUILDING EL 298-0 1 1 1 1 N 1 Not a pinch point. No action needed.COL N-Q, ROW 7.5-8.5 32 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Fire KSFs Lost or Impacted Pinch Recommendations Fire Zone Area Fire Zone Description DHR INV Point Category (from Table SSD/CA RAI 03-3) Suppression Detection PWR ?RX SFP RX SFP FBZR298N-2 2 REACTOR BUILDING EL 298-0 1 1 1 1 N 1 Not a pinch point. No action needed.COL N-Q, ROW 7.5-8.5 FBZR298S-1 1 REACTOR BUILDING EL 298-0 1 1 1 1 N 1 Not a pinch point. No action needed.COL K-L, ROW 7-8 FBZR298S-2 2 REACTOR BUILDING EL 298-0 1 1 1 1 N 1 Not a pinch point. No action needed.COL K-L, ROW 7-8 FBZR318N-1 1 REACTOR BUILDING EL 318-0 1 1 1 1 N 1 Not a pinch point. No action needed.COL M-Q, ROW 6-7 FBZR318N-2 2 REACTOR BUILDING EL 318-0 1 I I I I N 1 Not a pinch point. No action needed.COL M-Q, ROW 6-7 FBZR318S-1 1 REACTOR BUILDING EL 318-0 I I I I N 1 Not a pinch point. No action needed.COL K-M, ROW 6-7 FBZR318S-2 2 REACTOR BUILDING EL 318-0 1 I I I I N 1 Not a pinch point. No action needed.COL K-M, ROW 6-7 1_ ____FBZR340N-1 1 REACTOR BUILDING EL 340-0 1 I I I I N 1 Not a pinch point. No action needed.COL M-Q, ROW 6-7 FBZR34ON-2 2 REACTOR BUILDING EL 340-0 1 1 1 1 N 1 Not a pinch point. No action needed.COL M-Q, ROW 6-7 FBZR340S-1 1 REACTOR BUILDING EL 340-0 N N N N N N 1 Not a pinch point. No action needed.COL L-N, ROW 7-8 FBZR340S-2 2 REACTOR BUILDING EL 340-0 N N N N N N 1 Not a pinch point. No action needed.COL L-N, ROW 7-8 TURBINE BUILDING FIRE FBZT261N 5 BREAK ZONE NORTH EL 261-0 I I I I I N 1 Not a pinch point. No action needed.TURBINE BUILDING FIRE FBZT261S 5 BR NE SUTH E L L L L L Y 2 1B and/or 3B and/or 5 and/or 8 None None BREAK ZONE SOUTH EL 261-0 OG1 GENERAL FLOOR AREA EL I I I I I N 1 Not a pinch point. No action needed.232-0 OG2 5 GENERAL FLOOR AREA EL I I I I N 1 Not a pinch point. No action needed.247-0 OG3 5 GENERAL FLOOR AREA EL I L I L I Y 2 1A and/or 2A and/or 3A Yes Yes 261-0 33 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Fire FieZoeDecKSFs Lost or Impacted Pinch Recommendations Fireeoneecom m enPoi taCa egor Fire Zone Area Fire Zone Description DHR INV Point Category (from Table SSD/CA RAI 03-3) Suppression Detection RX SFP RX SFP R1 3 DRYWELL EL 237 -318 I I L I I Y 2 1B and/or 3B and/or 5 None None CTS PUMP ROOM AND R1A 1 GENERAL FLOOR AREA EAST I L L L I Y 2 1A and/or 2A and/or 3A Yes Yes EL 198-0 & 237-0 CTS PUMP ROOM, CS PUMP R1B 2 ROOM, GENERAL FLOOR L L L I I Y 2 1A and/or 2A and/or 3A Yes Yes AREA WEST EL 198-0 & 237-0 ACCESS STAIRWELL RIC 1 SOUTHEAST EL 237-0 & 261-0 I I I Not a pinch point. No action needed.CS PUMP ROOM AND PROTECTIVE CLOTHING R1D 1 CHANGE AREA EL 198-0 & 237- I I I I Not a pinch point. No action needed.0 R2A 1 GENERAL FLOOR AREA EAST L L L L I Y 2 1A and/or 2A and/or 3A Yes Yes EL 261-0 R2B 2 GENERAL FLOOR AREA WEST L L L I I Y 2 1A and/or 2A and/or 3A Yes Yes EL 261-0 R2C 2 SHUTDOWN COOLING ROOM I L L I I Y 2 1A and/or 3B and/or 5 Yes Yes EL 261-0 R2D 2 REACTOR BUILDING TRACK I I I I I N 1 Not a pinch point. No action needed.BAY EL 261-0 R3A 1 GENERAL FLOOR AREA EAST L L L L I Y 2 1A and/or 2A and/or 3A Yes Yes EL 281-0 R3B 2 GENERAL FLOOR AREA WEST I L L I I Y 2 1A and/or 2A and/or 3A Yes Yes EL 281-0 R4A 1 GENERAL FLOOR AREA EAST L L I L I Y 2 1A and/or 3B and/or 5 None Yes EL 298-0 R4B 2 GENERAL FLOOR AREA WEST L L I I I Y 2 1A and/or 2A and/or 3A Yes Yes EL 298-0 EMERGENCY CONDENSER R4C-1 1 ISOLATION VALVE ROOM EL I I I I I N 1 Not a pinch point. No action needed.298-0 34 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Fire KSFs Lost or Impacted PinchReomnais Fire Zone Fire Fire Zone Description DHR INV Point Category Recommendations Suppression Detection Area PW PontRateor (from Table SSD/CA RAI 03-3)RX SFP RX SFP PWR ?EMERGENCY CONDENSER R4C-2 2 ISOLATION VALVE ROOM EL I I I I I N 1 Not a pinch point. No action needed.298-0 R5A 1 GENERLFLOORAREAEST I L I L I Y 2 1A and/or 3B and/or 5 None Yes EL 318-0 R51 2 GENERAL FLOOR AREA WEST I I I I N 1 Not a pinch point. No action needed.EL 318-0 1 R6A 1 GENERAL FLOOR AREA EAST I I I I N 1 Not a pinch point. No action needed.EL 340-0 R6B 2 GENERAL FLOOR AREA WEST N N N N N N 1 Not a pinch point. No action needed.EL 340-01 RS1A 15 DRUM WASTE STORAGE N N N N N N 1 Not a pinch point. No action needed.VAULTS EL 252-0 RS1B 15 ELECTRICAL EQUIPMENT I I I I I N 1 Not a pinch point. No action needed.ROOM EL 252-0 GENERAL FLOOR AREA RSIC 15 SOUTH, DRUM STORAGE N N N N N N 1 Not a pinch point. No action needed.ROOM EL 252-0 RS2A 15 TRUCK LOADING AREA, N N N N N N 1 Not a pinch point. No action needed.NORTH EL 261-0 RS2B TRUCK LOADING AREA, WEST N N N N N N 1 Not a pinch point. No action needed.EL 261-0 RS2C 15 GENERAL FLOOR AREA EL N N N N N N 1 Not a pinch point. No action needed.261-0 RS2D) 15 RADWASTE CONTROL ROOM, N N N N N N 1 Not a pinch point. No action needed.WEST EL 261-0 RS2E 15 GENERAL FLOOR AREA, N N N N N N 1 Not a pinch point. No action needed.SOUTH EL 261-0 RS3A 15 GENERAL FLOOR AREA, N N N N N N 1 Not a pinch point. No action needed.WEST EL 281-0 RS4A 15 GENERAL FLOOR AREA, N N N N N N 1 Not a pinch point. No action needed.NORTHWEST EL 292-0 35 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)FireZone Fire FirZKSFs Lost or Impacted Pinch Recommendations Fire Zone Areaie Fire Zone Description DHR INV Point Category (from Table SSD/CA RAI 03-3) Suppression Detection RX SFP RX SFP RS5B 15 GENERAL FLOOR AREA, N N N N N N 1 Not a pinch point. No action needed.SOUTHWEST EL 292-01 S1 13 SCREENHOUSE EL 225-0 -I L I I I Y 2 1A and/or 3B and/or 5 None Yes 256-0 S2 14 DIESEL FIRE PUMP ROOM EL N N N N N N 1 Not a pinch point. No action needed.256-0 TURBINE T1 5 CONDENSER/HEATER BAY I I I L I Y 2 1A and/or 2A and/or 3A Yes Yes AREA EL 250-0 TURBINE BUILDING EL 240-261 TIA 5 MSIV ROOM & STEAM TUNNEL I I I I I N 1 Not a pinch point. No action needed.T2A 6 TURBINE BUILDING EL 250-0 L L L L L Y 2 1A and/or 2A and/or 3A Yes Yes T2B 7 TURBINE BUILDING SOUTH L L L L L Y 2 1A and/or 2A and/or 3A Yes Yes AND WEST EL 250-0 T2C TURBINE BUILDING OFFGAS I I I I I N 1 Not a pinch point. No action needed.TUNNEL EL 250-0 T2D 9 TURBINE BUILDING GENERAL L L L L L Y 2 1A and/or 2A and/or 3A Yes Yes AREA EAST EL 250-0 T2E 7 UPS BATTERY ROOM EL 250 N N N N N N 1 Not a pinch point. No action needed.GENERAL FLOOR AREA EAST T3A 5 OF MSIV ROOM AND FIRE L L L L L Y 2 1A and/or 2A and/or 3A Yes Yes ZONE T1 EL 261-318 GENERAL FLOOR AREA WEST T3B OF MSIV ROOM; ALSO SOUTH L L L L L Y 2 1Aand/or 2A and/or 3A Yes Yes AND WEST OF FIRE ZONE 1 EL 237-0 & 261-0 T4A 5 GENERAL FLOOR AREA EAST L L L L L Y 2 1A and/or 2A and/or 3A Yes Yes OF FIRE ZONE T1 EL 277-0 T4B 5 GENERAL FLOOR AREA WEST L L L L L Y 2 1A and/or 2A and/or 3A Yes Yes OF FIRE ZONE T1 EL 277-0 T4C HYDROGEN SEAL OIL UNIT N N N N N N 1 Not a pinch point. No action needed.ROOM EL 277-0 T4D 5 BATTERY ROOM EL 277 I I I I I N 1 Not a pinch point. No action needed.36 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)reZone Fire FieoKSFs Lost or Impacted Pinch Recommendations Fire Zone Area Fire Zone Description DHR INV Point Category (from Table SSDICA RAI 03-3) Suppression Detection RX SFP RX SFP T5A 5 GENERAL FLOOR AREA I L I L I Y 2 1A and/or 2A and/or 3A Yes Yes NORTH EL 291-0 1 1 T6A 5 GENERLFLOORAREA I L I L I Y 2 1A and/or 2A and/or 3A Yes Yes NORTH EL 305-6 T6B 5 TURBINELAYDOWNAREA I L I L I Y 2 1A and/or 5 None Yes EAST EL 300-0 T6C 5 GENERAL FLOOR AREA I I I I I N 1 Not a pinch point. No action needed.SOUTH EL 300-0 T6D 5 MECHANICAL STORAGE AREA N N N N N N 1 Not a pinch point. No action needed.EL 320-0 T7A 5 GENERAL FLOOR AREA I L I L I Y 2 1A and/or 3B and/or 5 None Yes SOUTH EL 320-0 GENERAL FLOOR AREA NORTH EL 333-0, GENERAL T8A 5 FLOOR AREA NORTH EL 351- I I I I 1 N 1 Not a pinch point. No action needed.0, GENERAL FLOOR AREA EAST EL 369 GENERAL FLOOR AREA WEST T8B 5 EL 369-0 I I I I I N 1 Not a pinch point. No action needed.GEEL A 369-0 250 WD1 15 GENERAL AREA El 225-0 & I I I I N 1 Not a pinch point. No action needed.229-0 WD2 15 GENERAL AREA EL 247-0 N N N N N N 1 Not a pinch point. No action needed.WD3A 15 GENERAL AREA EL 261-0 N N N N N N 1 Not a pinch point. No action needed.WD3B 15 RADWASTE CONTROL ROOM N N N N N N 1 Not a pinch point. No action needed.EL 261-0 WD3C 15 BALER ROOM EL 261-0 N N N N N N 1 Not a pinch point. No action needed.WD3D 15 DOW SOLIDIFICATION AREA N N N N N N 1 Not a pinch point. No action needed.EL 261-0 WD3E 15 TRUCK BAY EL 261-0 N N N N N N 1 Not a pinch point. No action needed.WD4 1 WASTE BUILDING VENTILATION AREA EL 277-0 N N N N N N 1 Not a pinch point. No action needed.37 of 47 ENCLOSURE1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Table SSD/CA RAI 03-3: List of Recommendations for Identified Pinch Points (from FAQ 07-0040)No. FAQ 07-0040 NMP1 Specific Recommendation for Outage Planning Recommendation Category 2 Fire Zones Considerations Limit hot work in this fire zone Outage planning considers periods 1A Lii o oki hsfr oe of increased vulnerability for Prohibition or limitation of hot during HRE conditions.

l h in thisefareizone.

workin irearea duinglimiting hot work in this fire zone.work in fire areas during periods of increased Outage planning considers periods vulnerability.

1B Prohibit hot work in this fire zone of increased vulnerability for during HREs. prohibiting hot work in this fire zone.Verify that the available fire detection systems located in the Detection systems should be 2A fire zone are functional.

Post verified to be functional; i.e., not firewatch in affected fire zones feragied t, etc.Verification of operable prior to entering HRE conditions if tagged out, etc.detection and /or suppression system(s) are impaired.in the vulnerable areas. Verify that the available fire suppression systems located in the area are functional.

Post firewatch Suppression systems should be 2B* in affected fire zones prior to verified to be functional; i.e., not entering HRE conditions if tagged out.system(s) are impaired.Limit transient combustible Outage planning considers 3A storage in this fire zone during limiting the hazard of combustible Prohibition or limitation of HRE conditions.

materials.

combustible materials in fire areas during periods of increased vulnerability.

Prohibit transient combustible Outage planning considers 3B storage in this fire zone during prohibiting the hazard of HRE conditions.

combustible materials.

Plant configuration changes Power can be removed from various Outage planning considers using 4* (e.g., removing power from components and equipment as part of alternate equipment and/or the equipment once it is placed in outage configuration line-ups prior to equipment's position whenever its desired position).

entering HRE conditions.

removing power.Provision of additional fire patrols at periodic intervals or Provide a firewatch (continuous or Outage planning considers the other appropriate periodic) in this fire area during HRE appropriate compensatory compensatory measures (such coditins measures required during periods as surveillance cameras) conditions.

of increased vulnerability.

during increased vulnerability.

Use of recovery actions to RActivities that may impact KSFs 6* e oftigateepotentialosss tof Recovery actions to restore at least one should be limited and strictly 6* mitigate potential losses of KSF success path can be taken. sol elmtdadsrcl KSFs. controlled to mitigate losses.38 of 47 ENCLOSURE1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)No. FAQ 07-0040 NMP1 Specific Recommendation for Outage Planning Recommendation Category 2 Fire Zones Considerations Identification and monitoring Outage planning considers the 7* in-situ ignition sources for hazards from the introduction of"fire precursors" (e.g., combustible materials and sources equipment temperatures).

of fire precursors.

Reschedule the work to a Activities in these fire zones should be Outage planning considers 8 period with lower risk or rescheduled to a period of non-HRE limiting work during periods of higher defense in depth (DID). conditions.

HRE conditions.

Control Rooms are constantly manned 9 N/A locations.

No other actions are required N/A during HRE conditions.

Existing controls on switchyard activities 10 N/A during HRE conditions are adequate and N/A will manage fire risk as well.* These recommendations are not used.Part b The NMP1 NPO pinch point analysis was developed in accordance with the guidance contained in FAQ 07-0040. The FAQ 07-0040 endorsed "Recommendations" utilized at NMP1 to reduce fire risk during HREs are identified in Part a of this RAI response.

The additional reduction in risk offered by the "Recommended" strategies provides additional assurance that fire risk is minimized in areas susceptible to a loss of one or more KSFs during plant HREs.As discussed in the response to Part a and depicted in Table SSD/CA RAI 03-2, additional actions (e.g., pre-fire rack-out, locally pinning of valves, isolation of air supplies) are not relied upon as a strategy to reduce fire risk during HREs, including the impact of fire-induced spurious operations (single or multiple).

The assessment of potential risk reduction options (including input from Operations personnel) concluded that the actual additional risk posed by fire during HREs is best controlled through the methods identified in Table SSD/CA RAI 03-2. Specifically, the NMP1 NPO strategy does not credit the following methods: " Recovery Actions -Reliance on recovery actions during an outage is difficult to characterize for feasibility due to the many variables that could exist, such as blockage of normal routes, scaffolding impact on lighting, equipment/material staging and movement, contract personnel contingent, unusual equipment line ups, etc. For this reason, recovery actions are viewed as less predictable with respect to reliability and uncertainty in comparison to the risk reduction options selected.* Configuration Changes -The use of limited configuration changes to address in a preemptive manner certain high consequence fire-induced failures, most notably spurious operations of key valves, was considered.

However, after discussions with Operations personnel it was concluded 39 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)that the reduction in operational flexibility to respond to a broader range of potential accidents and abnormal conditions outweighs the marginal improvement in risk reduction associated with fire-induced spurious operations.

With specific reference to the potential vulnerability of shutdown cooling isolation valves 38-01, 38-02, and 38-13 to fire-induced spurious closure, deliberately entering a TS required action was evaluated as undesirable when viewed from a broader perspective beyond just potential fire events.Thus, the recommendations contained in Table SSD/CA RAI 03-2 are considered the best options to augment existing procedures for managing shutdown risk, including risk from fire, during HREs.Part c As shown in the response to Part a and in Table SSD/CA RAI 03-2, NMP1 does not credit recovery actions as a strategy to reduce shutdown fire risk during HREs. The rationale for not employing recovery actions is provided in the response to Part b above.40 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Safe Shutdown / Circuit Analysis RAI 07 Based on a review of the updated final safety analysis report (UFSAR), switchgear other than motor control centers (MCCs) use 125 VDC power for control of the electrically-operated circuit breakers so the breakers may be operated if AC power is lost. Dualfeeds are provided to the DC control bus on each power board for added reliability, one each from either battery 11, 12 or 14.A generic concern in regards to the Fort Calhoun fire that occurred on June 7, 2011 (NRC Special Inspection Report, March 12, 2012, ADAMS Accession No. ML12072A128) involves 125 VDC circuits from both DC buses inside the same switchgear.

Both DC buses were impacted with "soft" grounds that remained after the fire had been isolated by removing power.With respect to the Fort Calhoun event, it appears that tile power boards at NMPJ have dual control power feeds. Describe if this issue has been considered.

Describe if there are any proposed plans to perform modifications or procedure changes to address this issue.Response to Safe Shutdown / Circuit Analysis RAI 07 General Description of the NMP1 125 VDC System The safety related 125 VDC system at NMP 1 consists of two physically separate and independent trains (Batteries 11 and 12). Each train includes one 125 VDC station battery, two parallel static battery chargers (one primary and the other a backup), and one DC power distribution board. The battery boards include the fuses and the fuse blocks required for distribution of 125V DC to various system loads. The augmented quality 125 VDC system consists of one 125 VDC station battery (Battery 14), a static battery charger, and one DC power distribution battery board.The 125 VDC batteries 11, 12, and 14 are part of NMP1 Safe Shutdown Equipment.

Battery 11 and the associated battery board are located in Fire Areas 17B and 16B, respectively.

Battery 12 and the associated battery board are located in Fire Areas 17A and 16A, respectively.

Battery 14 and the associated battery board are located in Fire Area 5.Ground Detection Design Features The 125 VDC electrical distribution trains are operated independently and ungrounded, and, as such, a single ground does not generate a fault current or disable the system. The system is equipped with ground detection devices to indicate the occurrence of the first ground which allows operators to locate and correct the first ground. NRC Information Notice (IN) 94-80 "Inadequate DC Ground Detection in Direct Current Distribution Systems," alerted licensees to the potential for operating with undetectable grounds in vital direct current (DC) distribution systems due to inadequate ground-detection equipment or inadequate ground-alarm setpoints, or both. The IN recommended that ground detectors be incorporated in the DC systems so that, if a single ground does occur, personnel are aware of the ground and can take immediate steps to clear the ground from the system. Failure to promptly eliminate a single ground could mask subsequent additional grounds. Multiple grounds could lead to unpredictable spurious operation of equipment, inoperable equipment, unanalyzed loads on batteries, or unanalyzed equipment failure modes.In response IN 94-80, the 125 VDC system ground detection scheme at NMP 1 was modified.41 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)A ground on the 125 VDC power system will be determined by the Ground Detection Relay and annunciated in the Control Room at panel A3 windows A3-4-2 for Battery 11, A3-4-3 for Battery 12, and A3-2-2 for Battery 14. Alarm Response Procedure Ni-ARP-A3 directs operators to Operating Procedure Nl-OP-47A (125 VDC System), Section H.8.0, which provides direction to locate and clear a ground within the 125 VDC system.System Loads Configuration For added reliability, dual DC feeds are provided to a number of Power Boards, Emergency Diesel Generators (EDGs) 102 and 103, and the Diesel Fire Pump (DFP). The dual feeds are selected via a mechanically operated knife switch located at each load and aligned to the normal DC feed. Table SSD/CA RAI 07-1 below provides a summary of loads with dual feeds.Table SSD/CA RAI 07-1: 125 VDC Electrical Distribution System Dual Feed Loads on Battery Boards 11, 12, and 14 Battery Board Load Battery Board Battery Board Battery Board#11 #12 #14 Motor Generator (MG) Set 167 Normal Alternate Breaker Control -Power Board 11 Normal Alternate Breaker Control -Power Board 12 Alternate Normal Breaker Control -Power Board 13 Alternate Normal Breaker Control -Power Board 14 Alternate Normal Breaker Control -Power Board 15 Alternate Normal Breaker Control -Power Board 16 Normal Alternate Breaker Control -Power Board 17 Alternate Normal Breaker Control -Power Board 18 Alternate Normal Breaker Control -Power Board 101 Normal Alternate Breaker Control -Power Board 102 Normal Alternate Breaker Control -Power Board 103 Alternate Normal DC Valve Board 11 Normal Alternate DC Valve Board 12 Alternate Normal Diesel Fire Pump Normal Alternate Hydrogen and Seal Oil System Annunciation Alternate Normal Stator Water System Annunciation Alternate Normal Emergency Diesel Generator 102 Starting and Control Normal Alternate Emergency Diesel Generator 103 Starting and Control Alternate Normal 42 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)NMP 1 Electrical Maintenance Procedures for Breakers and Switchiear As documented in the NRC Special Inspection Report 05000285/2011014 (Accession No.ML12072A128), the fire at Fort Calhoun Station occurred due to a high impedance connection which caused failure of a 480 VAC Breaker. The high impedance connection was caused by hardened grease on the secondary disconnects and dirty secondary contacts in GE model AKD-5 low voltage switchgear.

The root cause analysis determined that the electrical maintenance procedure associated with the low voltage switchgear was less than adequate, that preventive maintenance activities were inadequate to ensure proper cleaning of conductors, proper torquing of bolted conductor and bus bar connections, and that inspections for abnormal temperatures were inadequate.

At NMP1, procedure Ni-EPM-GEN-310 implements preventive maintenance on 4.16KV Switchgear, 600VAC Switchgear, 600 and 480 VAC Motor Control Centers (MCCs), and 125 VDC Battery Power Boards. Attachments 1 and 2 of this procedure include specific steps which address the issues identified in the Fort Calhoun root cause analysis, as follows:* When maintaining GE AKD-5 Load Master Switchgear, the breaker primary disconnecting studs and fingers are cleaned and greased with a thin coat of Mobil 28. Mobil 28 is selected based on its performance characteristics, which include resistance to friction oxidation (fretting) and hardening under various environmental conditions.

  • Inspect bolted connections, and torque any loose connections in accordance with specific torque requirements.

Procedure NI-EPM-GEN-151 for the inspection of TYPE AK-50 and ITE K-LINE breakers includes the following precautions and steps:* Prevent the mixing of Mobil 28 with previously used GE D50H15 or GE D50H47 lubricants since it may result in grease hardening and breaker failure.* Inspect main, intermediate, and arcing movable and stationary contacts for discoloration that may have been caused by overheating.

Procedure N I-EPM-GEN-182 for the inspection of MCCs includes the following precaution:

To prevent component insulation degradation, Trichloroethane (CRC Lectra Clean) based solvents shall not be used on the component insulating parts of the MCC or MCC Bucket. Denatured and isopropyl alcohol are acceptable substitutes for cleaning/degreasing of component insulated parts.THC-based solvents may be used for cleaning/degreasing of current carrying parts in metallic MCCs as well as on metallic MCC bucket components.

A review of past Condition Reports associated with breakers and breaker maintenance at Nine Mile Point was performed to determine if there has been any Condition Reports initiated due hardening of grease, dirty contacts, or loose connections.

The following provides a summary of this review: 43 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Condition Report Description 1994-000170 OE -Misadjustment between GE 4.16 KV circuit breakers and their associated cubicles 2000-003715 OE 11485, Cleaning of silver contacts caused silver to be removed 2001-001049 INPO SEN 218, Circuit breaker fault results in fire, LOOP, Reactor scram, and severe turbine damage 2001-004474 INPO SEN 221, Circuit Breaker to Bus Connector Faults Results in Reactor Scram. The corrective actions associated with this CR included revisions to Electrical Preventive Maintenance procedures NI -EPM-GEN-1 50 and NI -EPM-GEN-3 10 to incorporate steps/precautions to not use abrasive cleaner when cleaning silver plated contact surfaces.2003-001128 GE Service Information Letter 448, Rev. 1, Recommendation for lubrication of Type AK GE Breakers 2004-001184 Catastrophic Failure of MCCB 19A in 2NHS-MCCO10 due to phase-to-ground fault near line side connection.

A corrective action of this Condition Report implemented procedure revisions to include specific steps and directions for inspection of MCCB line side power wiring (procedures NI -EPM-GEN-1 82 and 310).The above Condition Reports and the associated corrective actions have resulted in procedure changes that address the potential causes of the fire event at Fort Calhoun.Differences Between NMP1 and Fort Calhoun In addition to lack of adequate preventive maintenance, the NRC Special Inspection Report 05000285/2011014 identified two addition contributing causes to the overall event at Fort Calhoun, as follows: Implementation of a plant modification in 2009 that replaced AK-50 480 V main and bus-tie breakers with Molded Case Square-D Masterpact circuit breaker/cradle assemblies and digital trip devices.The differences in form, fit, and function resulted in high resistance connections between the cradle assembly and bus stabs due to oxidation built-up caused by dissimilar metal (copper and silver) which contributed to the fire.There has been no modification implemented at NMP1 to replace breakers with the type identified above." Unlike Fort Calhoun, the NMP1 DC feeds to power boards, EDGs, and the DFP are equipped with fuses. These fuses function to effectively clear and isolate the affected battery board from an overcurrent condition caused by a hot short or multiple shorts to ground.44 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)Analysis of the Postulated Fire Scenario at NMP 1 A circuit analysis for the NMP1 NFPA 805 transition was performed in accordance with NEI 00-01, Revision 2, which has been endorsed by Regulatory Guide (RG) 1.189, Revision 2. In accordance with the guidance provided in NEI 00-01, Revision 2, evaluation of a potential "soft" ground (i.e.; a ground that does not result in sufficient fault current to cause the circuit protective/isolation device to open) is not required.

The following are excerpts from NEI 00-01, Section 3.5, with respect to ungrounded 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 electrical protective 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." Consider an individual, single short-to-ground on each conductor in each affected cable in a grounded circuit. Consider the combined effects of shorts-to-ground if conductors are located in the same multi-conductor cable in the primary circuit." For ungrounded circuits, two shorts-to-ground are required for the loss of control power to the individual circuit. The recommended approach either assumes or evaluates for a second short-to-ground causing a loss of control power in the components control circuit for ungrounded circuits." Additionally, either assume a second short-to-ground exists in an ungrounded circuit resulting in a loss of control power or evaluate for an actual fire-induced cable impact with the potential to cause the second short-to-ground in the fire area." Depending on the coordination characteristics between the protective device on the circuit and upstream circuits, the power supply to other circuits could be affected.

If multiple grounds can occur in a single fire area, they should be assumed to occur simultaneously unless justification to the contrary is provided.In summary, the concern with respect to a postulated short to ground on an ungrounded DC control circuit is multiple fire induced grounds that could result in a loss of control capability due to opening of the isolation devices.Similar to Fort Calhoun Station, NMP1 125 V DC battery boards 11, 12, and 14 provide redundant control power to a number of power boards, EDG 102 and 103 start and control circuits, and the DFP, as listed in Table SSD/CA RAI 07-1. A mechanically operated transfer switch allows the operators to manually re-align control power from the normal to the alternate 125 VDC battery board. The battery board loads listed in Table SSD/CA RAI 07-1 are located in Fire Areas 1, 2, 4, 5, 14, 19, 22, 23, and 24.The NMP1 Nuclear Safety Capability Assessment (NSCA) for the above Fire Areas did not address the potential for a "soft" ground on the 125 VDC system. Fundamental assumptions in the circuit analysis and fire area assessment are that: 1. A single short to ground will not affect the ability of the credited DC system to accomplish its intended safe shutdown function, and 45 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)2. Multiple hot shorts or shorts to ground will result in sufficient fault current as to cause actuation of the protective devices.A postulated fire within the Power Boards will potentially cause a single ground in the alternate 125 VDC supply. However, this single ground will not affect the safe shutdown function(s) of the credited 125 VDC bus. This is consistent with the conclusions of the NRC Special Inspection Report for the Fort Calhoun event, which states on page 26: "The team concluded that dc control power remained available to the safety-related 4160 VAC buses throughout the event, and the grounds on the dc buses would not have prevented the dc system from performing its safety function.

Because the system was normally ungrounded, a single ground on either the positive or negative bus of the system did not result in the loss of a circuit, but did indicate a degraded condition." Thus, the postulated "soft" ground on the credited 125 VDC electrical distribution system does not affect the assumptions in the NMP1 NSCA performed to support NFPA 805 transition with respect to separation requirements for redundant trains of the 125 VDC system. The analysis demonstrates that sufficient separation exists to ensure one train of the 125 VDC system remains free of fire damage. A postulated fire at each of the power board locations does not affect the capability to maintain battery charging to the unaffected train of 125 VDC and, as such, a minimal leakage current (i.e.; below the fuse opening) due to a "soft" ground would not affect battery capacity or charging capability.

Electrical Separation and Independence The safety related electrical distribution system at NMP1 is designed to provide two redundant and independent trains of control and power to safety related loads during and following anticipated transients and design basis accidents.

The design basis requirement also includes a criterion for limiting fire damage to one train of the electrical distribution system.To prevent paralleling the two trains of the safety related 125 VDC electrical distribution system, thereby losing train independence and redundancy, the following interlocks are provided: " The 125 VDC circuit breakers feeding computer MG Set 167 from DC battery boards 11 and 12 are key interlocked to prevent closing both breakers at the same time." The 125 VDC circuit breakers feeding DC valve boards 11 and 12 from 125 VDC battery boards 11 and 12 are mechanically interlocked to prevent closing both breakers at the same time.The 125 VDC system design and configuration meets the electrical separation requirements and single failure criterion and remains in compliance with the existing plant licensing basis based on the following: " The system is equipped with a ground detection circuit," Each control power feed is equipped with isolation devices which will effectively isolate the affected battery board from an overcurrent condition or hot short, and 46 of 47 ENCLOSURE 1 NINE MILE POINT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROPOSED ADOPTION OF NFPA 805 (2001 EDITION)* The dual feeds from the redundant 125 VDC power boards to each compartment are selected via a mechanically operated knife switch located in each power board to ensure electrical separation is maintained.

Conclusion The event at Fort Calhoun is considered Operating Experience (OE). Although the existing NRC guidance (RG 1.189, NUREG-6850) and industry guidance (NEI 001-01) do not require evaluation of a"soft" ground as part of the circuit analysis and fire area assessments performed for NFPA 805 transition, the Fort Calhoun fire scenario has been evaluated for applicability to NMP1. This evaluation has concluded that the occurrence of a fire caused by a lack of proper breaker preventive maintenance and the resulting consequences is not a likely fire scenario at NMP1. This is mainly due to differences in design configuration and maintenance activities at NMP 1. In addition, the fuses associated with each DC feed to the loads are sized to ensure that any short to ground faults are effectively isolated from the affected battery board.However, as documented in Information Notice 94-80, multiple grounds could lead to unpredictable spurious operation of equipment, inoperable equipment, and unanalyzed battery loads or equipment failure modes. To enhance operator knowledge and plant response to the potential for "soft" grounds, a change to post-fire safe shutdown procedure N1-SOP-21.1

("Fire In Plant") is being processed.

This change will alert the operators to the potential for a fire induced ground in the DC system following a confirmed fire in the plant, thereby enhancing reliability and defense-in-depth with respect to maintaining the availability of 125 VDC control power. No plant modifications or other procedure changes are deemed necessary to address this issue.47 of 47 ENCLOSURE 2 NINE MILE POINT UNIT 1 UPDATED RESPONSES TO PROBABILISTIC RISK ASSESSMENT RAI 05 AND RAI 08 Nine Mile Point Nuclear Station, LLC April 30, 2013 ENCLOSURE2 NINE MILE POINT UNIT 1 UPDATED RESPONSES TO PROBABILISTIC RISK ASSESSMENT RAI 05 AND RAI 08 By letter dated February 27, 2013, Nine Mile Point Nuclear Station, LLC (NMPNS) provided responses to requests for additional information documented in the NRC's letter dated January 3, 2013. In the February 27, 2013 letter, NMPNS committed to provide updates to the responses for Probabilistic Risk Assessment RAI 05 and Probabilistic Risk Assessment RAI 08 (if needed) to reflect the response to Safe Shutdown / Circuit Analysis RAI 01, in which the definition of the Nine Mile Point Unit 1 (NMP1) safe and stable condition has been revised. Each NRC RAI is repeated (in italics), followed by the updated NMPNS response.

Changes to the responses are identified by revisions bars drawn in the right margin.Probabilistic Risk Assessment RAI 05 Section 10 of NUREG/CR-6850, Supplement 1, states that a sensitivity analysis should be performed when using the fire ignition frequencies in the supplement instead of the fire ignition frequencies provided in Table 6-1 of NUREG/CR-6850.

Provide the sensitivity analysis of the impact on using the supplement I frequencies instead of the Table 6-1 frequencies on core damage frequency (CDF), large early release frequency (LERF), delta (4)CDF, and ALERF for all of those bins that are characterized by an alpha that is less than or equal to one. If the sensitivity analysis indicates that the change in risk acceptance guidelines would be exceeded using the values in Table 6-1, justify not meeting the guidelines.

Updated Response to Probabilistic Risk Assessment RAI 05 The NMP 1 Fire PRA uses the ignition frequencies from the latest guidance related to fire PRAs as given in Supplement 1 to NUREG/CR-6850.

Supplement 1 to NUREG/CR-6850 (Section 10.2) addresses the use of the ignition frequencies therein as follows: "The NRC accepts use of these revised fire bin ignition frequencies for fire PRAs conducted for NFPA-805 transition for best-/point-estimate calculations of fire risk (core damage frequency

[CDF]and large early release frequency

[LERF]), including delta-risk values from plant change evaluations, with the following provision.

The fire PRA, including plant change evaluations, must also evaluate the sensitivity of the risk and delta-risk results to evaluations performed using the current fire bin ignition frequencies in EPRI 1011989, NUREG/CR-6850, Chapter 6, "Fire Ignition Frequencies," Table 6-1, "Fire Frequency Bins and Generic Frequencies," and Appendix C, "Determination of Generic Fire Frequencies," Table C-3, "Generic Fire Ignition Frequency Model for U.S. Nuclear Power Plants." For those cases where the results from this sensitivity analysis indicate a change in the potential risk significance associated with elements of the fire PRA or plant change evaluations that affects the decisions being made (e.g., what is acceptable with the new frequencies from EPRI 1016735 might not be acceptable with the current applicable set from EPRI 1011989, NUREG/CR-6850), the licensee must address this situation by considering fire protection, or related, measures that can be taken to provide additional defense in-depth." With respect to the required sensitivity analysis, a footnote provides the following clarification: "The sensitivity analyses should be performed for a fire ignition frequency bin using the mean of the fire ignition frequency bins contained in NUREG/CR-6850.

Furthermore, sensitivity analyses only need to be performed for those bins characterized by an alpha from the EPRI 1016735 analysis that is less than or equal to 1. Note that an alpha value less than or equal to 1 is characteristic of a reverse-J shaped probability density function, i.e., the same shape as the non-informative prior distributions used in EPRI 1016735. This reverse-J shape is indicative of the large uncertainty in the bin fire 1 of 7 ENCLOSURE2 NINE MILE POINT UNIT 1 UPDATED RESPONSES TO PROBABILISTIC RISK ASSESSMENT RAI 05 AND RAI 08 frequency due to the sparsity of data for that bin, and therefore, the potential for significant changes should the post-2000 fire event data differ significantly from the 1991-2000 data. The required sensitivity analysis is, for the purpose of this interim solution, judged to provide an adequate indication of the effects on risk and delta-risk in such a case." Results of the Sensitivity Analysis Table PRA RAI 05-1 lists the ignition frequencies with alpha values < 1. Table PRA RAI 05-2 lists the risk results when the ignition frequencies from NUREG/CR-6850 are used, as well as the risk results as reported in the updated LAR Transition Report, which are based on Supplement 1 to NUREG/CR-6850 ignition frequencies.

Table PRA RAI 05-3 lists, at the fire area level, the risk results using ignition frequencies from NUREG/CR-6850 with alpha values < 1.An evaluation of the sensitivity results against Regulatory Guide 1.174 indicates that the delta risks by fire area in Table PRA RAI 05-3 using the NUREG/CR-6850 ignition frequencies meet the risk acceptance guidelines illustrated for Regions II and III of Figures 4 and 5 in Regulatory Guide 1.174 on an individual fire area basis. However, the total increase in risk associated with the implementation of NFPA 805 for the overall plant calculated by summing the risk increases exceeds the acceptance guidelines, as summarized in Table PRA RAI 05-2.Excluding fire risk, the plant risks associated with internal events, seismic, and high winds are estimated from Table 1 in the Fire Risk Evaluations (FRE) Report (NI -FRE-FOO 1, Revision 0) and are also shown in Table PRA RAI 05-2 below. Summing those risks with the fire risks gives the total plant CDF and LERF including fire and other risks (Table PRA RAI 05-2). Total CDF and LERF including non-fire risks remain below the critical levels of 10-4 for CDF and 10-5 for LERF. However, the ACDF and ALERF results exceed the delta risk guidelines (10-5 for CDF and 10-6 for LERF).Table PRA RAI 05-3 below lists the contribution to delta CDF and delta LERF for each fire area when the Fire PRA model is quantified using the frequencies from NUREG/CR-6850.

The results indicate that most of the contribution to delta CDF is generated by Fire Area 05 (-81%) and Fire Area 11 (- 1I%).These two areas are the top contributors to LERF as well. Consistent with the guidance in Section 10.2 of Supplement 1 to NUREG/CR-6850, fire protection, or related, measures that can be taken to provide defense in-depth for these three areas are discussed in the following paragraphs.

Justification for Not Meeting the Guidelines with the Higher Ignition Frequencies As suggested in Section 10.2 of Supplement 1 to NUREG/CR-6850, NMPNS has identified fire protection and related measures that provide additional defense-in-depth (DID), as justification for the sensitivity analysis results not meeting the delta risk guidelines.

These measures are presented in the Fire Risk Evaluation Report (Ni-FRE-FO01) and are summarized below for the fire areas that contribute the most to the calculated delta risk.Defense in Depth Measures for Fire Area 05 Fire Area 05 is relatively large covering most of the turbine building above elevation 261'. This fire area can be classified in two groups of fire zones. The first group gathers those fire zones where there is no installed automatic fire suppression system, or there is such a system but no credit is taken for it 2 of 7 ENCLOSURE2 NINE MILE POINT UNIT 1 UPDATED RESPONSES TO PROBABILISTIC RISK ASSESSMENT RAI 05 AND RAI 08 in the Fire PRA. This group consists of Fire Zones FBZT261S, OGI, OG2, OG3, TiA, T4B, T4D, T5A, T6A, T6B, T6C, T6D, T7A, T8A, and T8B.* There is no credit in the Fire PRA for the installed fire detection systems, automatic fire suppression systems, and manual suppression in Fire Zones FBZT261 S, T5A, T6C, T6D and T8A. The following systems are available in these zones and are credited for DID: Zone Detection Systems for DID Suppression Systems for DID FBZT261S DA-2161E, DA-2161M WP-2161 T5A D-2294, D-2304 SP-2314, SP-2324 T6C D-2385, D-2395, D-2395FL, D- WD-2395FL 2395PL T6D DA-2375 WP-2375 T8A D-2445, D-2485 SP-2465* In Fire Zone TiA, there are no installed fire detection systems or automatic fire suppression systems, and no credit is taken in the Fire PRA for manual suppression.

Manual fire suppression by the fire brigade is credited for DID.* In Fire Zones OGI, OG2, T4D, T6B, T7A and T8B, no credit is taken for the installed fire detection system and subsequent manual suppression (no automatic suppression system is installed in these fire zones). The following systems are available and are credited for DID: Zone Detection Systems for DID OGI D-7013 OG2 D-7013 T4D D-2194 T6B D-2355, D-2405 T7A DA-2425 T8B D-2485" In Fire Zones OG3, T4B, and T6A, no credit is taken for the installed automatic fire suppression system, but the Fire PRA credits the installed fire detection systems and subsequent manual suppression (manual suppression in T6A is credited only for the structural steel fire scenario).

The following systems are available and are credited for DID: Zone Suppression System(s) for DID OG3 SP-7053 T4B SP-2224, WP-2092 T6A SP-2314, C-2365 The second group of fire zones is made up of the balance of fire zones in the fire area; i.e.: TI, T3A, T3B, T4A, T4C, and FBZT261N.

For these fire zones, the Fire PRA takes credit for installed fire detection systems and automatic suppression systems, as well as manual suppression.

The local CO 2 fire suppression system is credited in TI, but with manual actuation only.3 of 7 ENCLOSURE 2 NINE MILE POINT UNIT 1 UPDATED RESPONSES TO PROBABILISTIC RISK ASSESSMENT RAI 05 AND RAI 08 Defense in Depth Measures for Fire Area 11 Fire Area II is equipped with a manual CO 2 system (C-303 1) credited for DID in Fire Zone C2. Fire Zone C2 is the Auxiliary Control Room (or Relay Room) located under the main control room. This system is not credited in the Fire PRA.Defense in Depth Measures Applicable for All Fire Areas The governing procedures for fire protection activities are GAP-INV-02, "Control of Material Storage Areas," and GAP-FPP-02, "Control of Hot Work." These procedures are not credited explicitly in the Fire PRA (i.e., the Fire PRA does not include failure probabilities to follow the requirements of these procedures) for postulating transient fires within Fire Area 05. The procedures are considered in the Fire PRA consistent with the guidelines in NUREG/CR-6850 for selecting the appropriate credit for prompt suppression and hotwork manual suppression curve for the appropriate scenarios and for determining the influence factors serving as weighting factors for transient fire ignition frequencies.

Consequently, the specific provisions of these procedures are credited for DID for: (1) controlling transient combustibles throughout the plant; and (2) assigning compensatory measures to maintenance activities that may temporarily change the plant configuration.

Effect of Response to Safe Shutdown / Circuit Analysis RAI 01 LAR Section 4.2.1.2 originally defined the NMP1 safe and stable condition as "the ability to maintain K~ff< 0.99 with a reactor coolant temperature at or below the requirements for hot shutdown and then subsequently cool down and maintain NMP1 in a cold shutdown condition." In the response to Safe Shutdown / Circuit Analysis RAI 01, the definition of the NMP1 safe and stable condition for the"At-Power" analysis has been revised to hot shutdown.

Analyses performed to support the response to Safe Shutdown / Circuit Analysis RAI 01 indicate an improvement to the delta risk numbers. In particular, the plant-level ACDF decreased from 1.52E-05/yr to 1.11E-05/yr, and the ALERF decreased from 1.71 E-06/yr to 1.29E-06/yr.

Table PRA RAI 05-1: Ignition Frequencies with Alpha Less than or Equal to 1 T. 1 NUREG/CR Frequency Supp. 1 UE/R Ratio: Supp. 1 Ignition Source (Location)

Supp. 1 Mean -6850 Mean NUREG/CR Bin Alpha Frequency Frequency

-6850 to (1 / y) (1 / y) -6850 to Supp.1 1 Batteries (Battery Room) 0.5 3.26E-04 7.5E-04 2.3 4 Main control board (Control Room) 1 8.24E-04 2.5E-03 3.0 9 Air Compressors (Plant-Wide) 0.5 4.65E-03 2.4E-03 0.5 11 Cable fires caused by welding and cutting 1 9.43E-04 2.OE-03 2.1 (Plant-Wide) 13 Dryers (Plant-Wide) 0.5 4.20E-04 2.6E-03 6.2 15.1 Electrical Cabinets Non-HEAF (Plant- 0.453 2.36E-02 4.5E-02 1.9 Wide)22 RPS MG sets (Plant-Wide) 0.92 9.33E-04 1.6E-03 1.7 31 Cable fires caused by welding and cutting 0.5 4.50E-04 1.6E-03 3.6 (Turbine Building)4 of 7 ENCLOSURE2 NINE MILE POINT UNIT 1 UPDATED RESPONSES TO PROBABILISTIC RISK ASSESSMENT RAI 05 AND RAI 08 Table PRA RAI 05-2: Sensitivity Study Results Result with Supplement I Sensitivity Result with Risk Measure Ignition Frequencies NUREG/CR-6850 Rkere (Ref. Updated LAR Table Ignition Frequencies

(/yr) W-3, N1-FRE-F001, Revision 1)CDF (fire) 2.06E-05 3.29E-05 CDF (other) 5.26E-06 5.26E-06 CDF (total) 2.59E-05 3.82E-05 LERF (fire) 2.23E-06 4.60E-06 LERF (other) 2.07E-06 2.07E-06 LERF (total) 4.30E-06 6.66E-06 ACDF 8.47E-06 1.11E-05 ALERF 7.05E-07 1.29E-06 Table PRA RAI 05-3: Delta Risks by Fire Area Fire Irea ACDF ACDF ALERF ALERF Fyr) Contribution

(/yr) Contribution 01 2.62E-07 2.37% 5.32E-08 4.12%02 1.08E-07 0.97% 3.20E-08 2.48%04 O.OOE+00 0.00% 1.60E-15 0.00%05 8.92E-06 80.69% 4.11E-07 31.83%06 1.26E-07 1.14% 1.27E-08 0.98%07 2.26E-07 2.04% 1.11E-08 0.86%09 2.32E-08 0.21% 1.48E-09 0.11%10 3.25E-08 0.29% 4.01E-09 0.31%II 1.26E-06 11.43% 7.01E-07 54.29%12 3.17E-11 0.00% 1.63E-09 0.13%13 4.34E-09 0.04% 3.87E-09 0.30%14 0.OOE+00 0.00% 0.OOE+00 0.00%15 1.29E-09 0.01% 6.OOE- 12 0.00%16A 0.OOE+00 0.00% 0.OOE+00 0.00%16B 0.OOE+00 0.00% 0.OOE+00 0.00%17A 0.OOE+00 0.00% 0.OOE+00 0.00%17B 0.00E+00 0.00% O.OOE+00 0.00%18 6.08E-10 0.01% 4.04E-10 0.03%19 8.84E-12 0.00% 1.49E-11 0.00%20 5.54E-11 0.00% 1.03E-11 0.00%21 5.82E-1 1 0.00% 1.04E-1 1 0.00%22 8.13E-08 0.74% 5.43E-08 4.21%23 6.50E-09 0.06% 4.34E-09 0.34%24 6.95E-10 0.01% 2.79E- 11 0.00%EXT N/A N/A N/A N/A Sum 1.11E-05 100.00% 1.29E-06 100.00%5 of 7 ENCLOSURE2 NINE MILE POINT UNIT 1 UPDATED RESPONSES TO PROBABILISTIC RISK ASSESSMENT RAI 05 AND RAI 08 Probabilistic Risk Assessment RAI 08 The transition report describes and justifies an initial coping time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, after which, actions are necessary to maintain safe and stable beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Provide a discussion of the actions necessary during and beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to maintain safe and stable conditions beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> such as refillingfluid tanks or re-aligning systems. Evaluate quantitatively or qualitatively the risk associated with these actions and equipment necessary to maintain safe and stable beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> given the post-fire scenarios during which they may be required.Updated Response to Probabilistic Risk Assessment RAI 08 The PRA model uses a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time for success criteria, similar to other PRAs and consistent with ASME/ANS RA-Sa-2009.

The plant must be in a safe stable state (e.g., hot shutdown condition) during this timeframe.

Decay heat levels are lower after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of success (safe stable state with inventory control and heat removal) in the PRA model and offsite resources and recoveries are available in case of any failures after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The probability of failures that may occur after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is considered negligible when other capabilities to recover are included.

Those support system dependencies in the PRA that are potentially sensitive to time have been evaluated.

The following summarizes these considerations: " Condensate Storage Tank: This tank supports reactor pressure vessel (RPV) makeup from the feedwater (high pressure coolant injection

-HPCI) system and control rod drive (CRD) pumps, and is a source of emergency condenser makeup. Considering RPV makeup without the emergency condensers, this tank is judged inadequate to last 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. With the emergency condensers and no loss of coolant accident (LOCA) condition, the tank may be adequate for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. There are 40,000 gallons of makeup water available to the condensate storage tanks via gravity feed from the condensate demineralizer water storage tank. Additional makeup would be required for a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> mission time. Fire water make up to the emergency condenser makeup tanks is available.

Damage repair procedure Nl-DRP-OPS-001 has instructions for supplying the demineralized water storage tank from city water, service water, or fire water. The demineralized water storage tank can then be drained to the condensate storage tanks by opening manual valve 57-31 located at Turbine Building elevation 305', column line J-1. The Fire PRA was also updated to model the recovery actions aimed at ensuring long-term EC makeup tank water supply via the diesel-driven fire pump and quantitatively evaluate their risk." 125V DC Power: Since emergency AC power is required, the batteries need only be available on demand to support emergency diesel generator (EDG) starting and other initial start loads. As long as the static charger and AC power are available after this battery demand, the batteries are not required in the long term. The batteries cannot supply DC loads for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without AC power support.* EDG Fuel Supply: At full load, one EDG consumes 228 gallons of fuel oil per hour. Each EDG has a 12,000 gallon fuel oil storage tank and a 400 gallon day tank. This would allow operation for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The fuel oil storage tanks can be cross connected to allow operation of one EDG at full load for 4 days. Therefore, the EDG fuel oil supply will last for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.* Room Cooling: The only areas of concern in the PRA are the two EDG areas (roof fans and the roll door in each EDG room). All other areas were judged to have slow heat up rates and/or maximum temperatures were sufficiently low.6 of 7 ENCLOSURE 2 NINE MILE POINT UNIT 1 UPDATED RESPONSES TO PROBABILISTIC RISK ASSESSMENT RAI 05 AND RAI 08 The risks associated with activities occurring after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> have been evaluated qualitatively and are considered to be negligible and, thus, acceptable.

LAR Section 4.2.1.2 originally defined the NMP 1 safe and stable condition as "the ability to maintain Kef< 0.99 with a reactor coolant temperature at or below the requirements for hot shutdown and then subsequently cool down and maintain NMP1 in a cold shutdown condition." In response to Safe Shutdown / Circuit Analysis RAI 01, the definition of the NMP1 safe and stable condition for the "At Power" analysis has been revised to hot shutdown.

This change led to the elimination of several VFDRs which pertained to cold shutdown, and also led to the creation of new VFDRs associated with long-term water supply to the EC makeup tanks. These new VFDRs were addressed by taking credit for recovery actions whose feasibility was evaluated and the risk quantitatively evaluated in the Fire PRA.7 of 7 ENCLOSURE 3 REVISIONS TO THE LAR TRANSITION REPORT WITH CHANGES HIGHLIGHTED The following are revisions to the Transition Report (included with the License Amendment Request (LAR) submitted by Nine Mile Point Nuclear Station, LLC (NMPNS) letter dated June 11, 2012)resulting from the responses to NRC requests for additional information (RAI) Safe Shutdown / Circuit Analysis RAI 01 and Safe Shutdown / Circuit Analysis RAI 03. The revised Transition Report pages, with the changes highlighted to facilitate their identification, are as noted below.* Sections 4.2 and 4.3 (Pages 14 through 29a)* Table 4-3 (Pages 54 through 60)* Attachment A (Pages A-42 through A-44)" Attachment B (Pages B-I through B-102)" Attachment F (Pages F-6 and F-7)" Attachment G (Pages G-1 through G-41)Nine Mile Point Nuclear Station, LLC April 30, 2013 REVISIONS TO TRANSITION REPORT SECTION 4.2, NUCLEAR SAFETY PERFORMANCE CRITERIA SECTION 4.3, NON-POWER OPERATIONAL MODES Pages 14 through 29a with changes highlighted.

Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements structures housing equipment required for nuclear plant operations are considered as"power block" structures.

These structures are listed in Attachment I and define the "power block" and "plant".4.2 Nuclear Safety Performance Criteria The Nuclear Safety Performance Criteria are established in Section 1.5 of NFPA 805.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" Establishing Recovery Actions" 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 4.3.4.2.1.1 Compliance with NFPA 805 Section 2.4.2 Overview of Process NFPA 805 Section 2.4.2 Nuclear Safety Capability Assessment states: "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 post-fire safe shutdown analysis (SSA) methodology against the guidance provided in NEI 00-01, Revision 2, Chapter 3,"Deterministic Methodology," as discussed in Appendix B-2 of NEI 04-02. NMP1 used the guidance provided in NEI 00-01, Revision 2 because it is endorsed as an NMI, pil213Pge1 I NMP1, April 2013 Page 14 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements acceptable methodology in NRC RG 1.205 and due to feedback received as a result of NRC requests for additional information on other post-pilot plant LARs.The methodology is depicted in Figure 4-2 and consisted of the following activities: " Each specific subsection of NFPA 805 Section 2.4.2 was correlated to the corresponding section of Chapter 3 of NEI 00-01, Revision 2. 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 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 those sections that do not align, an assessment was made to determine if the failure to maintain strict alignment with the guidance in NEI 00-01 could have adverse consequences.

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.

If the section has no adverse consequences, these sections of NEI 00-01 can be dispositioned without further review.In addition, a review of NEI 00-01, Revision 3 was conducted against the guidance from NEI 00-01, Revision 2. There were no gaps relative to MSOs identified.

The comparison of the NMP1 existing post-fire SSA to NEI 00-01 Chapter 3 (NEI 04-02 Table B-2) was performed and documented in EIR 51-9133191, Nine Mile Point Unit I -Nuclear Safety Capability Assessment.

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 from NEI 00-01, Revision 2, Chapter 3 (as supplemented by the gap analysis to Revision 3) directly or meets the intent of the endorsed guidance with adequate justification as documented in Attachment B.Referenced documents are planned as being retained as post-transition documents.

NEI 00-01, Revision 2, Chapter 3 contains guidance criteria concerning identifying required and important to SSD components.

These specific guidance criteria are not applicable to plants transitioning to NFPA 805; therefore, they were not addressed for NMPI.NMPI, April 2013 Pagel1 I A Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Step I Assemble Documentatlon LI pDeternnne and Document Step 2 Applicability of NEI 00-01 Sections FrApialFor Applicable NEI 00-01 Sections, Perform Compa idson of SSD Method vs. NEI 00-01 No No Yese N 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 only 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 Keff <0.99, 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 <0.99 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.NMPI, April 2013 Page 16 I I NMP1, April 2013 Page 16 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Results Demonstration of the Nuclear Safety Performance Criteria for safe and stable conditions was performed in two analyses." At-Power analysis for potential fires while in the Power Operating Conditions (Reactor mode switch is in "Startup" or "Run" position and the reactor is critical or criticality is possible due to control rod withdrawal) or Shutdown Condition

-Hot (Reactor mode switch is in "Shutdown" position and reactor coolant temperature is greater than 212°F), but not on shutdown cooling mode of decay heat removal.incl!udin*g .ta.tup and run. This analysis is discussed in Section 4.2.4." Non-Power analysis for potential fires while in Shutdown Condition

-Hot and lower operating conditions., which 'includoc Hot Shutdown and below.. This analysis is discussed in Section 4.3.Based on the EIR 51-9133191, Nine Mile Point Unit I -Nuclear Safety Capability Assessment, the NFPA 805 licensing basis for a safe and stable condition is defined as the ability to maintain Keff < 0.99 with a reactor coolant temperature at or below the requirements for hot shutdown.

and- subsoquontly cool down -and- Maintain PIAI120 in a cold ,hutdown condition.

The At-Power analysis includes a primary and alternate means of achieving and maintaining safe and stable conditions.

The primary means of achieving and maintaining Hot Shutdown (HSD) is via the Emergency Cooling system. Either of the two redundant Emergency Cooling decay heat removal loops can achieve and maintain HSD. The Emergency Cooling system operates by natural circulation where steam flows upward to the condenser(s) and returns as condensate to the Reactor Pressure Vessel (RPV). Decay heat is removed through the transfer of heat from the reactor coolant to the shell side water of the Emergency Condenser(s),.G...l.

which vents the developed steam to atmosphere.

Operation of either Emergency Cooling loop can sustain HSD conditions for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> without the need for makeup. from the Condncsato Sto.ago Tank (-ST). Upon achieving HSD conditions, the plant is able to maintain safe and stable operation for an extended period of time using the Emergency Cooling system. After 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the Emergency Condenser makeup tanks can be replenished as needed using the diesel driven fire pump (DFP), which draws water from Lake Ontario (effectively an infinite source). Periodic refueling of the DFP is accomplished in accordance with existing plant procedures using an on-site fuel source. Reactor coolant makeup is required after 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, assuming a nominal Technical Specification leakage rate of 25 gpm. Makeup is provided via the Control Rod Drive (CRD) system using one of the CRD pumps drawing suction from the CST. AC power is required to operate a CRD pump. The diesel driven fire pump may be aligned to provide primary makeup in the event no CRD pump is available.

The Emergency Cooling system can be initiated either manually or automatically.

The RPS instruments and logic that automatically initiate the Emergency Cooling system on high reactor pressure or low-low reactor level have been included in the analysis.Manual initiation of the Emergency Cooling system can be accomplished from either the Control Room or Remote Shutdown Panel (RSP) if Control Room abandonment is NMPI, April 2013 Page 17 I I NIWIPI, April 2013 Page 17 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements necessary, depending on the fire location.

AC power is not required to manually initiate Decay Heat Removal (DHR) via the Emergency Cooling system.In the event the primary means of decay heat removal during HSD method-is not available (i.e., Emergency Cooling system not available), the plant can be maintained in HSD by opening three Electromagnetic Relief Valves (ERVs) in the automatic depressudzation system (ADS) and blowing steam to the Torus to reduce pressure.When reactor pressure reaches approximately 365 psig, Core Spray (CS) may be utilized to provide core cooling. AC and DC electrical power are required for this method of decay heat removal.The ADS system can be initiated either manually or automatically.

The RPS instruments and logic automatically initiate the ADS system on a combination of low-low-low reactor level and high drywell pressure.The CS system can be initiated either manually or automatically.

The RPS instruments and logic automatically initiate the CS system on low-low reactor level or high drywell pressure.The instrumentation and logic circuitry that automatically initiates ADS and CS have been included in the analysis.

Manual initiation of the ERVs and CS system is accomplished from the Control Room. In the event spurious actuations of the ERVs NMPI, April 2013 Page 17a I I NMP1, April 2013 Page 17a I Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements take place due to a Control Room fire, CS can be initiated manually from outside the Control Room if Control Room abandonment is necessary.

The oif .aWhiviAng and maintaining Cold Shutdown (CSD) i6s via the S-hubt-d-OWA Cooling SYMStm (SDC). WAhen roactor presr isrduced to 120 psig and rnn_ r* a.rnG # ~ nm

  • e-n, ;- e-A'AGO +kn i +n a kn +rr r +;m^Ainn, *n t" k ~I inima~rRa[I*3f TkA Qflr' n ni4,nA ki., Danr Ei, inun inet f'Innn,q I e-en Mn M (RBCLC=) and Eimergoncy Sendaio Wiato W.OoSCpm nascainwt RBCLCG and ESW ~srqio oachioVo and maintain C-SD cnitions.

AC Gpowori roquirod to 8ntitoGD.Undor thsscnriractor coolan-t makeupis reud after R hoursm with an assumed Dove.' (CRD) system usng9 of the C-RD1 pumps draw~ing suctionn fr-m. t-ho CsT. AC-rrqu o or a iE6 pump. iT no iesel nHr pump may; aso he algedt proid makeup in the e':ent no CRD pump is, av.ailabo The alternate means of decay heat removal can be used to maintain safe and stable conditions until such time that the Shutdown Cooling (SDC) system is placed in service.in the event the pnimar'; CSD method is not av.ailable, the plant can be coolo-d dowAMn to-C.S.D using the CS syr.. m. CS is a two loop system. Operation of one loop is adequate to ensure core cooling.ahiove-CSD-.

When utilizing CS, the reactor vessel eventually floods to the point where the ERVs are passing fluid to the Torus rather than steam, in essence placing the RCS in recirculation through the Torus. During this process, decay heat is removed by operation of the Containment Spray (CTS) system in conjunction with the Containment Spray Raw Water (CTSRW) system. This mo'thod will bring th plant to CSG .Fully flooding the RPV negates the need for another system to provide inventory makeup. AC power is required to initiate this method.For either the primary or alternate means of achieving and maintaining safe and stable conditions, AC power is available from either the station Emergency Diesel Generators (EDGs) or offsite power. Actions required to achieve and initially maintain hot shutdown conditions can be performed by the minimum shift complement consisting of reactor operators, senior reactor operators, and non-licensed plant operators.

The EDGs can be refueled in accordance with existing plant procedures using an on-site fuel source (tanker truck), until such time that offsite power is restored.

Additional resources from the emergency response organization will be available to support EDG refueling activities.

LGIng TerM Safe and rStabl s ill baien;ntaiinl using either the preferred ot alIte-rna teA CSD m ethod. Wit AC h e~e availableM froinm either the stdationn Em-FeFgenc Diesel Generato (EDGs) Vfite power, C5[D nscan be maintained 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 OMAs 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 I NMPI, April 2013 Page 18 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements 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 transition pre-transition OMAs and to determine the population of post-transition recovery actions. This process is based on FAQ 07-0030 (ML1 10070485) and consists of the following steps:* Step 1: Clearly define the primary control station(s) and determine which pre-transition OMAs are taken at primary control station(s) (Activities that occur in the Main Control Room are not considered pre-transition OMAs). Activities that take place at primary control station(s) or in the Main Control Room are not recovery actions, by definition.

NMPI, April 2013 Page 18a I I NIVIPI, April 2013 Page 18a I Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements" Step 2: Determine the population of recovery actions that are required to resolve variances from deterministic requirements (VFDRs) (to meet the risk acceptance criteria or maintain a sufficient level of defense-in-depth)." Step 3: Evaluate the additional risk presented by the use of recovery actions required to demonstrate the availability of a success path." Step 4: Evaluate the feasibility of the recovery actions." Step 5: Evaluate the reliability of the recovery actions.Results The review results are documented in EIR 51-9156521, Recovery Action Review for Nine Mile Point Nuclear Power Station Unit I Transition to NFPA 805. Refer to Attachment G for the detailed evaluation process and summary of the results from the process.4.2.1.4 Evaluation of Multiple Spurious Operations Overview of Process NEI 04-02 suggests that a licensee submit a summary of its approach for addressing potential fire-induced MSOs for NRC review and approval.

As a minimum, NEI 04-02 suggests that the summary contain sufficient information relevant to methods, tools, and acceptance criteria used to enable the NRC to determine the acceptability of the licensee's methodology.

The methodology utilized to address MSOs for NMP1 is summarized below.As part of the NFPA 805 transition project, a review and evaluation of NMP1 susceptibility to fire-induced MSOs was performed.

The process was conducted in accordance with NEI 04-02 and RG 1.205, as supplemented by FAQ 07-0038 Revision 3 (ML1 10140242).

The BWR Generic MSO list in NEI 00-01, Revision 2, dated June 5, 2009 (including a gap analysis to the Revision 3 Generic MSO list) was utilized.The approach outlined in Figure 4-3 (based on Revision 3 from FAQ 07-0038) is one acceptable method to address fire-induced MSOs. This method used insights from the Fire PRA developed in support of transition to NFPA 805 and consists of the following: " Identifying potential MSOs of concern." Conducting an expert panel to assess plant specific vulnerabilities (e.g., per NEI 00-01, Rev. 1 Section F.4.2)." Updating the Fire PRA model and existing post-fire SSA to include the MSOs of concern." Evaluating for NFPA 805 compliance." Documenting results.This process is intended to support the transition to a new licensing basis. Post-transition changes would use the RI-PB change process. The post-transition change process for the 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.lNMPI, April 2013 Page 19 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Identification of new potential MSOs may be part of the plant change review process and/or inspection process).Identify Potential MVSOs of Concern*SSA Step 1 Generic List of MSOs Self Assessments PRA Insights Operating Experience Expert Panel Step 2 Identify and Document MSOs of Concern Update PRA model & NSCA (as appropriate) to include MSOs of concern Step 3

  • ID equipment* ID logical relationships
  • ID cables* ID cable routing Step 4 ComolianNo Pursue other resolution options Step 5 Document Results Figure 4-3 -Multiple Spurious Operations

-Transition Resolution Process (Based on FAQ 07-0038)Results Refer to Attachment F for the process used for NMP1 and the results from the process.4.2.2 Existing Engineering Equivalency Evaluation Transition Overview of Evaluation Process The EEEEs 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 EEEE review includes the following determinations:

0 The EEEE is not based solely on quantitative risk evaluations, I NMP1, April 2013 Page 20 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements" The EEEE is an appropriate use of an engineering equivalency evaluation," The EEEE is of appropriate quality," The standard license condition is met," The EEEE is technically adequate," The EEEE reflects the plant as-built condition, and" The basis for acceptability of the EEEE remains valid In 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, EEEEs 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' EEEEs, then the EEEE was referenced where required and a brief description of the evaluated condition was provided." If requesting specific NRC approval for 'adequate for the hazard' EEEEs, then the EEEE was referenced where required to demonstrate compliance and was included in Attachment L for NRC review and approval.In all cases, reliance on EEEEs to demonstrate compliance with NFPA 805 requirements is documented in the LAR.Results The review results for EEEEs are documented in EIR 51-9077683, NFPA 805 Fundamental Fire Protection Program and Design Elements Transition Review.In accordance with the guidance provided in RG 1.205, Regulatory Position 2.3.2, and NEI 04-02, as clarified by FAQ 07-0054, EEEEs used to demonstrate compliance with Chapters 3 and 4 of NFPA 805 are referenced in Attachments A and C as appropriate.

None of the transitioning EEEEs require NRC approval.4.2.3 Licensing Action Transition Overview of Evaluation Process The existing licensing actions (exemptions

/ safety evaluations) review 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.Results Attachment K contains the detailed results of the Licensing Action Review.The following licensing actions will be transitioned into the NFPA 805 fire protection program as previously approved (NFPA 805 Section 2.2.7). These licensing actions are considered compliant under 10 CFR 50.48(c).lNMP1, April 2013 Page 21 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements

  • None The following licensing actions are no longer necessary and will not be transitioned into the NFPA 805 fire protection program: " An exemption from the requirements of Section III.G.2 of Appendix R for the battery board rooms (FA 16A and FA 16B), since their boundary walls do not provide the required 3-hour rated barriers." An exemption from the requirements of Section III.G.2 of Appendix R for the battery rooms (FA 17A and FA 177B), since their boundary walls do not provide the required 3-hour rated barriers." An exemption from the requirements of Section III.G of Appendix R for the control room (FA 11), since the control room ceiling does not have a 3-hour rating from the control room side due to unprotected structural steel members." An exemption from the requirements of Section Ill.G.2 of Appendix R for the wall between the reactor building and the turbine building above elevation 340' (FA 1, FA 2, and FA 5), since the wall is not a 3-hour rated barrier." An exemption from the requirements of Section III.G.2 of Appendix R for the fire break zone separating FA 1 and FA 2 in the reactor building upper level (elevation 340'), since the wall is not a 3-hour rated barrier.These exemptions are no longer required because the subject boundaries have been demonstrated adequate for the hazard in an EEEE.Since the exemptions are either compliant with 10 CFR 50.48(c) or no longer necessary, in accordance with the requirements of 10 CFR 50.48(c)(3)(i), NMPNS requests that the exemptions listed in Attachment K be rescinded as part of the LAR process. See Attachment 0, Orders and Exemptions.

4.2.4 Fire Area Transition Overview of Evaluation Process The Fire Area Transition (NEI 04-02 Table B-3) was performed using the methodology contained in NEI 04-02 and FAQ 07-0054. The methodology for performing the Fire Area Transition, depicted in Figure 4-4, is outlined as follows: 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." Assessed accomplishment of nuclear safety performance goals. Documented the method of accomplishment, in summary level form, for the fire area." 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 documented the results of the review. See Section 4.2.3.I NMP1, April 2013 Page 22 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements" Performed existing engineering equivalency evaluation reviews. Performed a review of existing engineering equivalency evaluations (or created new evaluations) documenting the basis for acceptability.

See Section 4.2.2." Performed a review of pre-transition OMAs 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.Step 3 -VFDR identification, characterization and resolution considerations.

Identified variances from the deterministic requirements of NFPA 805, Section 4.2.3.Documented variances as either a separation issue or a degraded fire protection system or feature. Developed VFDR problem statements to support resolution.

Step 4 -Performance-Based evaluations (Fire Modeling or Fire Risk Evaluations).

See Section 4.5.2 for additional information.

Step 5 -Final Disposition." Documented final disposition of the VFDRs in Attachment C (NEI 04-02 Table B-3)." For recovery action compliance strategies, ensured the manual action feasibility analysis of the required recovery actions was completed.

Note: If a recovery action cannot meet the feasibility requirements established per NEI 04-02, then alternate means of compliance were considered." Documented the post transition NFPA 805 Chapter 4 compliance basis.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.NMPI, April 2013 Page 23 I I NMP1, April 2013 Page 23 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Identify INITIAL Variances From Determinisic Requirements of NFPA 805 § 4.2.3 (6-3 Table)Document Final Disposition of VFDR Compliance options include: Accept As Is Require FP systemslfeatures Require Recovery Action Require Programmatic Enhancements Require Plant Modifications (B-3 Table)Figure 4-4 -Summary of Fire Area Review[Based on FAQ 07-0054 Revision 1]NMPI, April 2013 Page 24 I I NMP1, April 2013 Page 24 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements 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.NEI 04-02 Table B-3 includes the following summary level information for each fire area: " Regulatory Basis -NFPA 805 post-transition regulatory bases." Performance Goal Summary -An overview of the method of accomplishment of each of the performance criteria in NFPA 805 Section 1.5." Reference Documents

-Specific references to Nuclear Safety Capability Assessment Documents." Licensing Actions -Specific references to safety evaluations that will remain part of the post-transition licensing basis. A brief description of the condition and the basis for acceptability of the licensing action are provided.

In addition, summaries of Fire Risk Evaluations performed for variances from the deterministic requirements are also provided." EEEE -Specific references to EEEE 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 are provided." VFDRs -Specific variances from the deterministic requirements of NFPA 805 Section 4.2.3. Refer to Section 4.5.2 for a discussion of the performance-based approach.4.3 Non-Power Operational Modes 4.3.1 Overview of Evaluation Process NMP1 implemented the process outlined in NEI 04-02 and FAQ 07-0040, "Non-Power Operations Clarifications." The goal (as depicted in Figure 4-5) is to ensure that contingency plans are established when the plant is in a Non-Power Operational (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: " Review of the existing Outage Management Processes" Identification of Equipment/Cables:

o Review of plant systems to determine success paths that support each of the defense-in-depth Key Safety Functions (KSFs), and o Identification of cables required for the selected components and determination of their routing." Perform Fire Area Assessments (identify pinch points -plant locations where a single fire may damage all success paths of a KSF)." Manage pinch-points associated with fire-induced vulnerabilities during the outage.!NMPI, April 2013 Page 25 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements The process is depicted in Figures 4-5 and 4-6. The results are presented in Section 4.3.2.Figure 4-5 Review POSs, KSFs, Equipment, and Cables, and Identify Pinch Points NMPI, April 2013 Page 26 I I NIVIP1, April 2013 Page 26 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements 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 Figure 4-6 Manage Pinch Points NMPI, April 2013 Page 27 IlNMP1, April 2013 Page 27 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements

4.3.2 Results

of the Evaluation Process Based on FAQ 07-0040, the Plant Operating States (POSs) considered for equipment and cable selection are defined in EIR 51-9171174, Nine Mile Point 1, Non-Power Operations KSF Equipment List. The methodology for determination of KSFs, success paths, components required to achieve the success paths and their associated cabling arewere defined in EIR 51-9137629, Nine Mile Point 1 -Nuclear Power Station -NFPA 805 Transition

-Non Power Operations Component Pinch Point Analysis.

These documents provide the component selection information for the POSs included in the analysis (HSD, CSD, Refueling and Defueled conditions).

Systems and components were identified to support the following KSFs:* Reactor Vessel Decay Heat Removal Capability

  • Spent Fuel Pool Cooling* Reactor Vessel Inventory Control* Spent Fuel Pool Inventory Control* Electrical Power Availability
  • Reactivity Control KSFs are evaluated in each fire zone. Each KSF has one or more paths that satisfy the specific function.

The process for selection and treatment of components is consistent with the methodology in the NSCA. Where new components are required to support KSFs, the components are included in NPO equipment list and cable identification is performed using the same methodology as that employed for the NSCA. Where NSCA equipment is also relied upon in the NPO analysis but the NPO functional requirements differed from that in the NSCA, additional reviews are performed to ensure comprehensive cable selection.

Inherent in the process is identification of components potentially vulnerable to single and multiple spurious operation concerns during NPO.Note that the Reactivity Control KSF is not included in the NPO analysis because it is administratively controlled in accordance with procedure NIP-OUT-01.:

DHR for both the R-e-mctor and the Spent Fuel Pool (SFP), ... ..,,tr.y Control for- bth theVeosel and the SF12, and Po wo r avai lability.

Each m1ay have one or moIre 15SM pathr, that satisfy thatoifcKF No effort iswas made to eliminate or reduce fire impact by circuit analysis; therefore, a conservative estimate of damage is provided, including hot-short induced spurious operation of equipment.

By assuming that a single fire impacts any and all components in a fire zone, (whether the individual component or its associated cables are physically located within the fire zone), the assumption is made that the entire contents of the fire zone are lost.EIR 51-9137629 contains the fire zone evaluations comprising the 'KSF pinch point'analysis.

If a component that is part of a particular KSF flew-path is impacted, it is assumed that the KSF path is lost. -However, there are normallymay-be one or more other flew-paths within the particular KSF that are not impacted; therefore, the KSF is not considered lost and does not constitute a pinch point. Only when all paths for a particular KSF are impacted-, is the KSF itself considered lost and identified is-as a pinch point.NMPI, April 2013 Page 28 I I NMPI, April 2013 Page 28 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements The NPO analysis results (EIR 51-9137629 and EIR 51-9171174 ) categorize each KSF (in each fire zone), as either 'T", "L" or "N" as follows: T "I" (Impacted):

At least one of the KSF paths associated with a giventhe KSF is affected, i.e., a component of a specific the-KSF path or any of the component's required cablesits aSSOciatod cables within the fire zone are impacted, whereby that path can no longer be assured of being functional.

However, at least one other KSF path for the within that-KSF remains is-still-available." "L" (Lost): All available success paths for a given KSF are impacted." "N" (None): No impacts to the KSF are identified."Pinch Points" are were-thenidentified (on a fire zone basis), based on the complete loss of a KSF. In accordance with FAQ 07-0040, any evaluated area in which all of the credited success paths for a given KSF are lost is considered a KSF pinch point. Each KSF for all Fire Zones is evaluated and documented in the pinch point analysis.

The Fire Zone is labeled as AP,-"N" if in the pinch POint ,olumn indicates that no KSFs arewere lost. in th;s firAe zone. The Fire Zone is labeled with aA "Y' in this "w'-mn ifddioates4hat-if one or more KSFs arewere lost, thereby identifying that the Fire Zone contains one or more pinch points.in this fire Zone and therefeo-e, constitutes, a pinch NUMARC 91-06 discusses the development of outage plans and schedules.

A key element of that process is to ensure the KSFs perform as needed during the various outage evolutions.

During outage planning, the NPO Fire Zone Assessment is reviewed to identify areas of single-point KSF vulnerability during HRE to develop needed contingency plans/actions.

Depending upon the significance of the damage for those areas, combinations of the following options to reduce fire risk are considered at a minimum: I NMPI, April 2013 Page 28a I!NMP1, April 2013 Page 28a I'A Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements" Prohibition or limitation of hot work in fire areas during periods of increased vulnerability;" Verification of functional fire detection and/or suppression in the vulnerable areas;" Prohibition or limitation of combustible materials in fire areas during periods of increased vulnerability;" Plant configuration changes (e.g., removing power from equipment once it is placed in its desired position);" Provision of additional fire patrols at periodic intervals or other appropriate compensatory measures (such as surveillance cameras) during increased vulnerability;" Use of recovery actions to mitigate losses of Key Safety Functions;" Reschedule the work to a period with lower risk or higher DID;" Identification and monitoring in-situ ignition sources for fire precursors.

In addition, KSF equipment removed from service during the HREs is evaluated.

The evaluation is based on KSF equipment availability and NPO Fire Zone Assessment for any necessary contingency plans/actions.

The recommendations for reducing risk can be applied as appropriate during the implementation phase (as part of the technical document and procedure review). NMP1 strategies for reducing NPO fire risk do not rely on Recovery Actions or Pre-Fire Actions (i.e., Plant Configuration Changes) as strategies for reducing NPO fire risNote that roc .'"or; acti.n. wor. not. uod- to FiiAtolmno firomd, ucod failuroc.See Attachment D for more complete details. Based on incorporation of the recommendations from the KSF pinch point evaluations into appropriate plant procedures prior to implementation of the NFPA 805 fire protection program, the performance goals for NPO modes are fulfilled and the requirements of NFPA 805 will be met. See Implementation Items in Table S-2 of Attachment S.4.4 Radioactive Release Performance Criteria 4.4.1 Overview of Evaluation Process The review of the fire protection program against NFPA 805 requirements for fire suppression related radioactive release was performed using the methodology contained in EIR 51-9085686, NMP-1 NFPA 805 Radiological Release Transition Review. The methodology consists of the following: " Screened the fire zones based on the potential for the presence of contaminated materials during all plant operating modes, including full power and non-power conditions.

The screening process considered input from radiation protection personnel and review of the NMP1 fire pre-plans.

The evaluation focused on radioactive release to any unrestricted area due to firefighting activities." Reviewed fire pre-plans and fire brigade training materials 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 NMPI, April 2013 Page 29 I I NMPI, April 2013 Page 29 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements 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 NMPI, April 2013 Page 29a I I NMP1, April 2013 Page 29a I REVISIONS TO TRANSITION REPORT TABLE 4-3,

SUMMARY

OF NFPA 805 COMPLIANCE BASIS AND REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES Pages 54 through 60 with changes highlighted.

Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features NFPA 805 Required Required Required Fire Fire Suppression Detection Protection Required Fire Protection Feature Area Fire Zone Description Regulatory System System Feature and System Details 1 Basis (S, E, R, D)2 (S, E, R, D)2 (S, E, R, D)2 1 Reactor Building East EL 198-0 thru EL 340-0 4.2.4.23 FBZR237N FBZR261N FBZR281N FBZR281S FBZR298N FBZR298S FBZR318N FBZR318S FBZR340N FBZR340S Reactor Building EL 237-0 COL N-Q, ROW 8-9 Reactor Building EL 261-0 COL N-Q, ROW 8-9 Reactor Building EL 281-0 COL M-Q, ROW 6-7 Reactor Building EL 281-0 COL K-L, ROW 7-8 Reactor Building EL 298-0 COL N-Q, ROW 7.5-8.5 Reactor Building EL 298-0 COL K-L, ROW 7-8 Reactor Building EL 318-0 COL M-Q, ROW 6-7 Reactor Building EL 318-0 COL K-M, ROW 6-7 Reactor Building EL 340-0 COL M-Q, ROW 6-7 Reactor Building EL 340-0 COL L-N, ROW 7-8 S,R S,R S,R S,R S,R S,R S,R S,R None None S,R S,R S,R S,R S,R S,R S,R S,R D D None None None None None None None None None None Water Pre-Action Sprinkler, Detection Water Pre-Action Sprinkler, Detection Water Pre-Action Sprinkler, Detection Water Pre-Action Sprinkler, Detection Water Pre-Action Sprinkler, Detection Water Pre-Action Sprinkler, Detection Water Pre-Action Sprinkler, Detection Water Pre-Action Sprinkler, Detection Detection Detection 1 R1A CTS Pump Room And General Floor Area East EL 198-0 D R None Water Pre-Action Sprinkler, Detection& 237-0 1 RiC Access Stairwell Southeast EL 237-0 & 261-0 None D None Detection 1 R1D CS Pump Room And Protective Clothing Change Area EL D D None Water Pre-Action Sprinkler, Detection 198-0 & 237-0 1 R2A General Floor Area East EL 261-0 D R None Water Pre-Action Sprinkler, Detection 1 R3A General Floor Area East EL 281-0 None RD None Detection 1 R4A General Floor Area East EL 298-0 None D None Detection Halon Suppression System, 1 R4C Emergency Condenser Isolation Valve Room EL 298-0 D D None Dtcon Detection 1 R5A General Floor Area East EL 318-0 None D None Detection I R6A General Floor Area East EL 340-0 None D None Detection 2 Reactor Building West EL 198-0 thru EL 340-0 4.2.4.23 2 FBZR237N Reactor Building EL 237-0 COL N-Q, ROW 8-9 S,R S,R None Water Pre-Action Sprinkler, Detection 2 FBZR261N Reactor Building EL 261-0 COL N-Q, ROW 8-9 S,R S,R None Water Pre-Action Sprinkler, Detection 2 FBZR281N Reactor Building EL 281-0 COL M-Q, ROW 6-7 S,R S,R None Water Pre-Action Sprinkler, Detection I NMPI, April 2013 Page 54 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features Fire NFPA 805 Required Required Required Fire Fire Suppression Detection Protection Required Fire Protection Feature Area Fire Zone Description Regulatory System System Feature and System Details 1 Basis (S, E, R, D)2 (S, E, R, D)2 (S, E, R, D)2 2 FBZR281S Reactor Building EL 281-0 COL K-L, ROW 7-8 S,R S,R None Water Pre-Action Sprinkler, Detection 2 FBZR298N Reactor Building EL 298-0 COL N-Q, ROW 7.5-8.5 S,R S,R None Water Pre-Action Sprinkler, Detection 2 FBZR298S Reactor Building EL 298-0 COL K-L, ROW 7-8 S,R S,R None Water Pre-Action Sprinkler, Detection 2 FBZR318N Reactor Building EL 318-0 COL M-Q, ROW 6-7 S,R S,R None Water Pre-Action Sprinkler, Detection 2 FBZR318S Reactor Building EL 318-0 COL K-M, ROW 6-7 S,R S,R None Water Pre-Action Sprinkler, Detection 2 FBZR340N Reactor Building EL 340-0 COL M-Q, ROW 6-7 None D None Detection 2 FBZR340S Reactor Building EL 340-0 COL L-N, ROW 7-8 None D None Detection 2 RIB CTS Pump Room, CS Pump Room, General Floor Area None R None Detection West EL 198-0 & 237-0 2 R2B General Floor Area West EL 261-0 D R None Water Pre-Action Sprinkler, Detection 2 R2C Shutdown Cooling Room EL 261-0 None D None Detection 2 R2D Reactor Building Track Bay EL 261-0 D None None Dry Pipe System 2 R3B General Floor Area West EL 281-0 None E,R None Detection 2 R4B General Floor Area West EL 298-0 None D None Detection 2 R4C Emergency Condenser Isolation Valve Room EL 298-0 D D None Halon Suppression System, Detection 2 R5B General Floor Area West EL 318-0 D D None Water Pre-Action Sprinkler, Detection 2 R6B General Floor Area West EL 340-0 None D None Detection 3 Drywell EL 237-0 thru 318-0 4.2.3.1 3 R1 Drywell EL 237 -318 None None None 4 Foam Room EL 261-0 4.2.4.23 4 AB1F Foam Room EL 261-0 None RD None Detection 5 Turbine Building EL 240-0 thru 369-0 4.2.4.23 5 FBZT261N Turbine Building Fire Break Zone North EL 261-0 E,R E,R None Water Pre-Action Sprinkler, Detection 5 FBZT261S Turbine Building Fire Break Zone South EL 261-0 E,RD E,RD None Water Pre-Action Sprinkler, Detection 5 OG1 General Floor Area EL 232-0 None E,D None Detection I NMPI, April 2013 Page 55 I!NMP1, April 2013 Page 55 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features NFPA 805 Required Required Required Fire Fire Suppression Detection Protection Required Fire Protection Feature Aire Fire Zone Description Regulatory System System Feature and System Details'Basis (S, E, R, D)f (S, E, R, D)2 (S, E, R, D)O 5 OG2 General Floor Area EL 247-0 None E,D None Detection 5 OG3 General Floor Area EL 261-0 E,D ER None Wet Pipe System, Detection 5 T1 Turbine Condenser/Heater Bay Area EL 250-0 E,R E,R None Deluge System, CO 2 required for risk, Detection 5 TIA Turbine Building EL 240-261 MSIV Room & Steam Tunnel None None None General Floor Area East of MSIV Room and Fire Zone T1 Water Pre-Action Sprinkler, Wet and 5 T3A EL 261-318 E,R,D E,R None Dry Pipe System, CO required for DID, Detection Water Pre-Action Sprinkler, CO 2 5 T3B General Floor Area West of MSIV Room; Also South And E,R,D E,R NeeS required for DID, Detection, Promat-H West Of Fire Zone 1 EL 237-0 & 261-0 enclosure for the HVAC duct Water Pro-Action Sprinkler, Wet Pipe 5 T4A General Floor Area East Of Fire Zone T1 EL 277-0 E,R E,R None and Deluge Sytem, D etion and Deluge System, Detection Water Pre-Action Sprinkler, Wet Pipe 5 T4B General Floor Area West Of Fire Zone T1 EL 277-0 E,D E,R None Sytem Detection System, Detection 5 T4C Hydrogen Seal Oil Unit Room EL 277-0 E,R,D E,R None Deluge System, Co 2 required for DID, Detection 5 T4D Battery Room EL 277 None E,D None Detection 5 T5A General Floor Area North EL 291-0 E,D E,D None Wet Pipe System, Detection 5 T6A General Floor Area North EL 305-6 ED E,R None Wet Pipe System, CO 2 required for DID, Detection 5 T6B Turbine Laydown Area East EL 300-0 None E,D None Detection 5 T6C General Floor Area South EL 300-0 E,D E,D None Deluge System, Detection 5 T6D Mechanical Storage Area EL 300-0 E,D E,D None Water Pre-Action Sprinkler, Detection 5 T7A General Floor Area South EL 320-0 None E,D None Detection T8A General Floor Area North EL 333-0, General Floor Area E,D E,D None Wet Pipe System, Detection North EL 351-0, General Floor Area East EL 369 5 T8B General Floor Area West EL 369-0 None E,D None Detection T .... EL 20 42" 6 T2A Turbine Building EL 250-0 R E,R None Water Pre-Action Sprinkler, Detection NMPI, April 2013 PageSe I I NMPI, April 2013 Page 56 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features NFPA 805 Required Required Required Fire Fire Suppression Detection Protection Required Fire Protection Feature Area Fire Z Desption Reulatory System System Feature and System Details'Basis (S, E, R, D)2 (S, E, R, D) (S, E, R, D)2 7 Tu 7 T2B Turbine Building South And West EL 250-0 R R None Water Pre-Action Sprinkler, Detection 7 T2E UPS Battery Room EL 250 None D None Detection 9 T2C Turbine Building Offgas Tunnel EL 250-0 None D None Detection 9 T2D Turbine Building General Area East EL 250-0 R R None Water Pre-Action Sprinkler, Detection 10~~Wae Pre-cmo Sprinkler, RoCO 5.04A.10 C1 Cable Spreading Room EL 250-0 E,R,DWater Pre-Action Sprinkler, 02 required for DID, Detection Conro Cmplx L 61- AidFEL27-04.2A 11 C2 Auxiliary Control Room, Computer Room 261-0 R,D ER NoneS Halon Suppression System, C02 required for DID, Detection, Promat-H enclosure for the HVAC duct I 11 C3 Control Room EL 277-0 None E,R None Detection 12 Amfbd uitE 5 4.42" 12 AB1A Records Storage Area EL 250-0 None D None Detection 12 AB1B SAS Equipment Area EL 252-0 D D None Halon Suppression System, Detection 12 ABIC CPU Equipment Area EL 252-0 D D None Halon Suppression System, Detection 12 ABID General Area EL 250-0 D D None Wet Pipe System, Detection 12 ABlE Locker Area, Lunch Room, Offices EL 261-0 D D None Wet Pipe System, Detection 12 AB2A Access Passageway EL 248-0 D None None Wet Pipe System 12 AB2B Technical Support Area EL 248-0 D D None Wet Pipe System, Detection 12 AB2C Radiation Records Area EL 248-0 D D None Wet Pipe System, Detection 12 AB2D Warehouse Area EL 248-0 D None None Wet Pipe System 12 AB3A Warehouse Area EL 261-0 D None None Wet Pipe System NMPI, April 2013 Page 57 I NMP1, April 2013 Page 57 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features NFPA 805 Required Required Required Fire Fire Suppression Detection Protection Required Fire Protection Feature Area Fire Zone Description Regulatory System System Feature and System Details'Basis (S. E, R, D)2 (S, E, R, D)2 (S, E, R, D)2 12 AB3B Oil Storage Room EL 261-0 D None None Wet Pipe System 12 AB3C Storeroom Truck Dock EL 261-0 D None None Dry Pipe System 12 AB3D Electrical/Mechanical Shop Area, Office Areas, Locker OR OR None Wet Pipe System, Detection Rooms EL 261-0 12 AB3E Telephone Room 1 EL 261 -0 D D None Halon Suppression System, Detection 12 AB3F Telephone Room 2 EL 261-0 D D None Halon Suppression System, Detection 12 AB4A General Office Area EL 277-0 D D None Wet Pipe System, Detection 12 AB4B File Room EL 277-0 D D None Water Pre-Action Sprinkler, Detection 12 AB4C Records Processing Area EL 277-0 R R None Water Pre-Action Sprinkler, Detection 12 AB4D General Office Area EL 277-0 R R None Wet Pipe System, Detection 12 AB5 Penthouse Ventilation Room EL 290-0 D D None Deluge System, Detection 13 S1 Screenhouse EL 225-0 -256-0 SD R None Dry Pipe System, Detection 14 Disl iePm o L2104.Z4js Wet Pipe System required for 14 S2 Diesel Fire Pump Room EL 256-0 SD SD None Chapter 3, Section 3.9.4, complianceD.y

.-"pe -System- , Detection 15 RS1A Drum Waste Storage Vaults EL 252-0 None None None 15 RS1B Electrical Equipment Room EL 252-0 D D None Halon Suppression System, Detection 15 RSIC General Floor Area South, Drum Storage Room EL 252-0 None D None Detection 15 RS2A Truck Loading Area, North EL 261-0 D None None Dry Pipe System 15 RS2B Truck Loading Area, West EL 261-0 D None None Dry Pipe System 15 RS2C General Floor Area EL 261-0 D D None Dry-pipe System, Detection I NMPI, April 2013 PageS8 I I NMPI, April 2013 Page 58 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features NFPA 805 Required Required Required Fire irea Fire Zone Description Regulatory Suppression Detection Protection Required Fire Protection Feature FArea Fr oeDsrpinRgltr System System )2 Feature and System Details'Basis (S, E, R, D)2 (S, E, R, D)2 (S. E, R, D)2 Halon Suppression System, 15 RS2D Radwaste Control Room, West EL 261-0 D D None Deton Detection 15 RS2E General Floor Area, South EL 261-0 None None None 15 RS3A General Floor Area, West EL 281-0 None D None Detection 15 RS4A General Floor Area, Northwest EL 292-0 D D None Deluge System, Detection 15 RS5B General Floor Area, Southwest EL 292-0 None D None Detection 15 WD1 General Area EL 225-0 & 229-0 None D None Detection 15 WD2 General Area EL 247-0 None D None Detection 15 WD3A General Area EL 261-0 D D None Water Pre-Action Sprinkler, Detection 15 WD3B Radwaste Control Room EL 261-0 None D None Detection 15 WD3C Baler Room EL 261-0 D D None Dry Pipe System, Detection 15 WD3D Dow Solidification Area EL 261-0 D D None Dry Pipe System, Detection 15 WD3E Truck Bay EL 261-0 E,D E,D None Dry Pipe System, Detection 15 WD4 Waste Building Ventilation Area EL 277-0 None D None Detection 16A BIA Battery Board Room 12 EL 261-0 None E,D None Detection 168~I IalyBadRo I i I EL 2iii42 16B BIB Battery Board Room 11 EL 261-0 None E,D None Detection 1IA Batter Ro I E I 17A B2A Battery Room 12 EL 277-0 None E,D None Detection 1713Ba~ey Rom II E 277ZII42Ae 17B B2B Battery Room 11 EL 277-0 None E,D None Detection 18 D3 EDG 102 Control Cable Missile Enclosure EL 271-0 None E,D NoneS Detection, Promat-H enclosure NMPI, April 2013 Page 59 I I NMP1, April 2013 Page 59 1 Constellation Energy Nuclear Group 4.0 Compliance with NFPA 805 Requirements Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features NFPA 805 Required Required Required Fire Fire Suppression Detection Protection Required Fire Protection Feature Area Fire Zone Description Regulatory System System Feature and System Details 1 Basis (S, E, R, D) (S, E, R, D)2 (S, E, R, D)2 19 Emergency Diesel Generator 103 Foundation Room 4.2.4.23 EL 250-0 and Diesel Generator Room El 261-0 19 D1A EDG 103 Foundation Room EL 250-0 E,D E,D None Water Pre-Action Sprinkler, Detection 19 D2A EDG 103 Room EL 261-0 R E,R None C02 System, Detection 20 Diesel Generator Enclosed Cableway EL 250-0 4.2.4.23 20 D1C EDG 103 Cable Routing Area EL 250-0 D D None Water Pre-Action Sprinkler, Detection 21 Below Powerboards 102 & 103 EL 250-0 4.2.4.23 21 D1D Room Below PB's 102 & 103 EL 250-0 D D None Water Pre-Action Sprinkler, Detection 22 Emergency Diesel Generator 102 Foundation Room 4.2.4.23 EL 250-0 and Diesel Generator Room EL 261-0 22 D1B EDG 102 Foundation Room EL 250-0 E,D E,D None Water Pre-Action Sprinkler, Detection 22 D2B EDG 102 Room EL 261-0 R E,R None C02 System, Detection 23 Power Board 102 Room EL 261-0 4.2.4.23 23 D2C Power Board 102 Room EL 261-0 D E,R None C02 System, Detection 24 Power Board 103 Room EL 261 4.2.4.23 24 D2D Power Board 103 Room EL 261-0 D E,R None C02 System, Detection EXT External to Plant 4.2.3.13 EXT EXT External to Plant E E None Deluge System, Detection Notes: 1. Refer to Attachment C for each area and additional information

2. NR -Not Required; S -Required for Separation; E -Required for Engineering Evaluation; R -Required for Risk; D -Required for Defense-in-Depth
3. Compliance includes reliance on simplifying deterministic assumptions NMPI, April 2013 Page 60 I!NMPI, April 2013 Page 60 1 REVISIONS TO TRANSITION REPORT ATTACHMENT A NEI 04-02 TABLE B-l, TRANSITION OF FUNDAMENTAL FIRE PROTECTION PROGRAM & DESIGN ELEMENTS Pages A-42 through A-44 with changes highlighted.

Attachment A -NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Constellation Energy Nuclear Group a.a.-I u I I itsl rt: pruL;tmUviI wdtt~r supply system shall be dedicated for fire protection use only.%.,Um pllest WitlI Clarification based on Exception No. I III dUUILIUII LU lilt! PFULUL.UII UbU, 1'4 I-OL.-U 10 dIIU LII1 UFSAR describe the following non-fire protection, nuclear safety / emergency uses of the fire protection water supply system: Exception No. 1: Fire protection water supply systems shall be permitted to be used to provide backup to nuclear safety systems, provided the fire protection water supply systems are designed and maintained to deliver the combined fire and nuclear safety flow demands for the duration specified by the applicable analysis.Exception No. 2: Fire protection water storage can be provided by plant systems serving other functions, provided the storage has a dedicated capacity capable of providing the maximum fire protection demand for the specified duration as determined in this section.a. Provide a source of make-up water to the emergency condenser make-up tanks.b. Provide an emergency source of water for containment and reactor vessel flooding.c. Provide an emergency source of make-up water to the spent fuel pool.d. Provide a back-up water source for the emergency service water system.e. Provide a back-up water source for the diesel generator cooling water system.The following additional considerations apply to non-fire protection uses described above: a. Use of the electric motor-driven or diesel engine-driven fire pump as a source of emergency make-up to the emergency condenser make-up tanks would not be required for a minimum of 8-hours4A hae-wg after depletion of the emergency condenser inventory, assuming worst case conditions in which the CST inventories are unavailable due to fire damage to the CST transfer system. If the CST transfer system is available, emergency condenser makeup using a fire pump is not required for a minimum of 48-hours.

aRd CST in'-ntorioc.

Thorof-rm, In either case, concurrent fire protection use is unlikely.UFSAR Sec. X.N, Appendix 10A, Rev.21, Sec. 2.5, pg. 1OA-29 UFSAR Sec. X.N, Appendix 10B, Rev.21, Appendix D, Sec. 1.c, pg. 1OB-220 Calculation

$13.1-100F005, "Diesel Fire Pump / Reactor Vessel Flooding," Rev. 0, All Calculation S13.1-100F006," Pressure Drop Calculation, NMP2 Main Fire Pumps Supply to NMP1 Fire Water Distribution System," Rev. 0, All Calculation 13.1-100F007, "Hydraulic Analysis of Diesel Fire Pump Supply to ESW #11 and Emergency Diesel Cooling Water Systems," Rev. 0, All NMPI, April 2013 Page A-42 I I NMPI, April 2013 Page A-42 I Constellation Energy Nuclear Group Attachment A -NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements ,bhaptei. renceand Compliance Statement Compliance Bases Reference Documents Reqgu irementslGUidance 3.5.16 (cont.) (cont.)b. Use of the electric motor-driven or diesel engine-driven fire pump for emergency containment and reactor vessel flooding would only be required in the unlikely event that all other means of core injection are lost. On this basis, concurrent fire protection use is unlikely.c. Use of the electric motor-driven or diesel engine-driven fire pump as a source of emergency spent fuel pool inventory would only be required upon loss of the Fuel Pool Make-up System, condensate transfer via hoses, and demineralized water with hoses from refueling service connections.

On this basis, concurrent fire protection use is unlikely.d. and e. Use of the diesel engine-driven fire pump as a source of ESW and EDG cooling water would only be required upon occurrence of a fire in the Screenhouse that could disable all other Screenhouse pumps.Clarification:

The required flow rate and pressure demand for each of the non-fire protection, nuclear safety / emergency uses of the fire protection water supply system described above can be supplied by use of one (1) of the redundant fire pumps. Therefore, in the unlikely event that concurrent fire protection and non-fire protection use is required, either the remaining Unit 1 fire pump, or the Unit 2 fire protection water supply system will be available.

To ensure adequate water supply for fire suppression activities concurrent with other uses, a modification for a cross-connection will be installed.

See modifications in Attachment S.NMI pi 01 aeA4 I NIVIPI, April 2013 Page A-43 I Attachment A -NEI 04-02 Table 13-1 Transition of Fundamental Fire Protection Program & Design Elements Constellation Energy Nuclear Group 3.5.16 (cont.) (cont.)Additionally, since the source of the Unit I and Unit 2 fire protection water supply is Lake Ontario, the water available to supply fire protection and/or non-fire protection demands is not a concern.3.6 Standpipe and Hose N/A N/A -Section title, no technical requirements.

See Stations.

sub-sections for specific compliance statements and references.

3.6.1 For all power block Complies with use of Standpipe and hose systems are provided for all NI-SD-018, Rev. 05, Sec. 2.2.5 buildings, Class III standpipe EEEE power block structures.

Per N1-SD-018, standpipe and hose systems shall be risers are located at various points throughout Nine UFSAR Sec. X.N, Appendix 10A, Rev.installed in accordance with Mile Point Unit I to serve hose stations.

The 21, Sec. 2.5.3.4, Table 1.2.2, NFPA 14, Standard for the standpipes/hose stations are so spaced as to permit pgs. I OA-46, 1OA-47, 1OA-86 Installation of Standpipe, hose stream coverage of all points in the buildings Private Hydrant, and Hose including primary containment.

Hose stations are FPEE 0-98-003, "Acceptable Use of Systems. equipped with 100 feet of 1-1/2 inch hose with Aluminum Fire Hose Couplings," Rev.adjustable spray nozzles. Hose stream coverage is in 0 accordance with NFPA 14 Class III systems.EIR 51-9077284-000, "NMP-1 Code Reviews," Sec. 4.0, Appendix F 3.6.2 A capability shall be Complies Based on the fire protection water supply design $13.1-100F002, -Fire Protection provided to ensure an adequate information contained in $13.1-100F002, the NMP1 Water Supply," Rev. 02 water flow rate and nozzle fire protection water supply system is capable of pressure for all hose stations.

providing adequate flow and pressure for all hose UFSAR Sec. X.N, Appendix 1OA, Rev.This capability includes the stations and exceeds NFPA 14-1963 design 21, Sec. 2.5.3.4, Table 1.2.2, provision of hose station requirements.

pgs. 1OA-46, 1OA-47, 1OA-86 pressure reducers where necessary for the safety of plant Pressure reduction devices are not installed for hose industrial fire brigade members stations.

This has been deemed acceptable because and off-site fire department fire hoses connected to the standpipe system are personnel.

intended for use exclusively by trained fire brigade I personnel.

I NMPI, April 2013 Page A-44 I I NMP1, April 2013 Page A-" I REVISIONS TO TRANSITION REPORT ATTACHMENT B NEI 04-02 TABLE B-2, NUCLEAR SAFETY CAPABILITY ASSESSMENT

-METHODOLOGY REVIEW Pages B-1 through B-102 with changes highlighted.

Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review B. NEI 04-02 Table B-2 -Nuclear Safety Capability Assessment

-Methodology Review 101 Pages Attached NMPI, April 2013 Page B-I I NMP1, April 2013 Page B-1 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 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 or result in the mal-operation of those components needed to meet the nuclear safety criteria shall be included.Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref 3 Deterministic Methodology NEI 00-01 Guidance This section discusses a generic deterministic methodology and criteria that licensees can use to perform a post-fire safe shutdown analysis to address regulatory requirements.

For a complete understanding of the deterministic requirements, work this section in combination with the information in Appendix C, High/Low Pressure Interfaces, Appendix D, Alternative and Dedicated Shutdown Requirements, Appendix E, Acceptance Criteria for Operator Manual Actions and repairs, and Appendix H, Hot Shutdown versus Important to Safe Shutdown Components.

To resolve the industry issue related to MSOs, refer to Section 4, Appendix B, Appendix F and Appendix G. The plant specific analysis approved by NRC is reflected in the plant's licensing basis. The methodology described in this section is an acceptable method of performing a post-fire safe shutdown analysis.

This methodology is depicted in Figure 3-1. Other methods acceptable to NRC may also be used. Regardless of the method selected by an individual licensee, the criteria and assumptions provided in this guidance document may apply. The methodology described in Section 3 is based on a computer database oriented approach, which is utilized by several licensees to model Appendix R data relationships.

This guidance document, however, does not require the use of a computer database oriented approach.Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis This document addresses the comparison of the deterministic methodology used for the existing the Nine Mile Point 1 (NMP1) Safe Shutdown Analysis and the requirements of 10CFR50 Appendix R, Sections IlI.G.1, III.G.2 and Ill.G.3 against the compliance requirements and criteria specified in NFPA 805. The subsequent sections determine the extent the analysis meets the requirements as described in NEI 04-02. This documents a line by line review and comparison against the methodology and criteria provided in Chapter 3 of NEI 00-01, Revision 2. The deterministic methodology described in Section 3 and Figure 3-1 of NEI 00-01 was utilized as documented in the following Table B-2 sections.

The NMP1 safe shutdown methodology utilizes a computer oriented database to model data relationships for systems, components, and cables used to comply with the requirements for post fire safe shutdown.

This review of modifications, procedural controls, repair procedures and previously approved configurations and boundaries demonstrates that the safe shutdown analysis generally meets the Nuclear Safety Performance Criteria including the information provided in Appendix B of NFPA 805 related to circuit criteria and Multiple Spurious Operations (MSOs).Reference Document EIR 51-9133191, NSCA, Section 9.0 NMPI, April 2013 Page B-2 I I NMP1, April 2013 Page B-2 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.1 Safe Shutdown Systems and Path Development NEI 00-01 Guidance This section discusses the identification of systems 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 III.G.1 .a requires that the capability to achieve and maintain hot shutdown be free of fire damage. Appendix R Section Ill.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 />. This section provides some guidance on classifying components as either required or important to SSD circuit components.

It also provides some guidance on the tools available for mitigating the effects of fire-induced circuit failures to each of these classes of equipment.

For a more detailed discussion of the topic of required and important to SSD components refer to Appendix H.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 required 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 required for 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 (as defined in NRC Information Notice 84-09)Support systems* Electrical power and control systems* Component Cooling systems* Component Lubrication 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.The components required to perform these functions are classified as required for hot shutdown components.

These components are necessary and sufficient to perform the required safe shutdown functions assuming that fire-induced impacts to other plant equipment/cables do not occur. Since fire-induced impacts to other plant equipment/cables can occur in the fire condition, these impacts must also be addressed.

The components not necessary to complete the required safe shutdown functions, but which could be impacted by the fire and cause a subsequent impact to the required safe shutdown components are classified as either required for hot shutdown or important to SSD components.

Depending on the classification of the components, the tools available for mitigating the effects of fire-induced damage vary. The available tools are generally discussed in this section and in detail in Appendix H. The classification of a component or its power or control circuits may vary from fire area to fire area. Therefore, the required safe shutdown path for any given fire area is comprised of required for hot shutdown components and important to SSD components.

The distinction and classification for each required safe shutdown path for each fire area should be discernible in the post-fire safe shutdown analysis.Generic Letter 81-12 specifies consideration of associated circuits of concern with the potential for spurious equipment operation and/or loss of power source, and the common enclosure failures.

As described above, spurious operations/actuations can affect the accomplishment of the required safe shutdown functions listed above. Typical examples of the effects of the spurious operations of concern are the following:

NMPI, April 2013 Page B-3 I!NMPI, April 2013 Page B-3 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review* A loss of reactor pressure vessel/reactor coolant inventory in excess of the safe shutdown makeup capability.

To address the issue of multiple spurious operations, Section 4 provides a Resolution Methodology for developing a Plant Specific List of MSOs for evaluation.

Appendix B provides the circuit failure criteria applicable to the evaluation of the Plant Specific list of MSOs.Common power source and common enclosure concerns could also affect the safe shutdown path and must be addressed.

In addition to the tools described for components classified as required for hot shutdown, fire-induced impacts to cables and components classified as important to SSD may be mitigated using some additional tools. For important to SSD component failures, operator manual actions, fire modeling and/or a focused-scope fire PRA may be used to mitigate the impact. (If the use of a Focused-Scope Fire PRAs is not permitted in the Plants Current License Basis, then, a License Amendment Request (LAR) will be necessary to use the Focused-Scope Fire PRA).Applicability Comments Applicable None Alignment Statement Aligns Alignment Basis To achieve post-fire safe shutdown, the Safe Shutdown Systems, their functions, and components required to support the safe shutdown functions were identified.

P&IDs and Electrical drawings were marked up and annotated to select equipment and specific flow paths for each system required to support safe shutdown.

This information was populated into a computer database to provide a database oriented approach to model Appendix R data relationships.

Safe shutdown systems, components, and cables were selected using the criteria and assumptions of NEI 00-01, Rev. 2, Sections 3.1.1 and 3.1.2 to satisfy each safe shutdown function performance goals, including process monitoring and support systems for each path. NEI 00-01 was used as the guidance in developing the Safe Shutdown Equipment List and the Safe Shutdown success paths. Safe Shutdown paths were designed based on the combination of systems in the respective fire area.The ability to achieve post-fire Safe Shutdown (SSD) is assured by having at least one safe shutdown path of the required systems, structures, and components to remain free of fire damage. This assurance that the safe shutdown equipment is available supports the required performance goals identified in the guidance, maintains the integrity of the fuel, reactor pressure vessel and primary containment.

The SSD path used to achieve post-fire safe shutdown is comprised of SSD systems and components that remain free of fire damage.Reference Document EIR 51-9133191, NSCA, Sections 4.0, 5.0, 5.1, 5.2, 8.6.2, and 8.6.3 HNP RAI 3-6, RAI 3-9, RAI 3-10, and RAI 3-11, NRC Requests for Additional Information dated August 6, 2009 (ML092170715)

Closure of National Fire Protection Association 805 Transition Program Frequently Asked Question Number 06-0006 (ML070030117)

ONS RAI 3-35, NRC Request for Additional Information dated November 18, 2009 (ML092920347)

NMPI, April 2013 Page B-4 I I NMP1, April 2013 Page B-4 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.1.1 Criteria/Assumptions NEI 00-01 Guidance The following criteria and assumptions should be considered, as applicable, when identifying systems available and necessary to perform the required safe shutdown functions and combining these systems into safe shutdown paths. This list provides recognized examples of criteria/assumptions but should not be considered an all-inclusive list. The final set of criteria/assumptions should be based on regulatory requirements and the performance criteria for post-fire safe shutdown for each plant.Applicability Applicable Alignment Statement Not Required Comments None Alignment Basis This paragraph provides introductory information and contains no specific requirements.

Discussion is provided in subsequent sub-sections.

Reference Document NMPI, April 2013 Page B-5 I I NMP1, April 2013 Page B-5 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.1.1.1 Safe Shutdown Paths For BWRs NEI 00-01 Guidance[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 III.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.Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis The primary means of achieving and maintaining hot shutdown following a fire coincident with a Loss of Offsite Power (LOOP) is via the Emergency Condenser (EC) system. The EC system consists of two redundant emergency cooling loops; each loop is capable of independently accomplishing hot shutdown.

Therefore, this option provides two redundant paths for obtaining hot shutdown.The EC system operates by natural circulation.

Steam flows from the vessel through the EC tubes. Condensate returns to the vessel through a reactor recirculation loop. Boiling of water in the secondary side of the ECs, which is vented to the atmosphere, provides the necessary cooling.In the event the preferred shutdown method is not available, the plant can be shut down by opening three Electromatic Relief Valves (ERVs) and discharging steam to the Torus to reduce pressure.

When reactor pressure reaches approximately 365psig, Core Spray (CS) may be initiated.

CS is a two loop system. Operation of one LOOP is adequate to achieve shutdown.

Eventually, the reactor vessel floods to the point where the ERVs are passing fluid to the Torus rather than steam, in essence placing the Reactor Coolant System in recirculation through the Torus. During this process, decay heat is removed by operation of the Containment Spray System in conjunction with the Containment Spray Raw Water system. This shutdown method will bring the plant directly to CSD. Fully flooding the Reactor Pressure Vessel negates the need for another system to provide inventory makeup. AC power is required to initiate this shutdown method.Reference Document EIR 51-9133191, NSCA, Sections 5.1 and 5.2 NMPI, April 2013 Page B-6 I NMP1, April 2013 Page B-6 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.1.1.2 SRVs and LPCI/CS NEI 00-01 Guidance[BWR] GE Report GE-NE-T43-00002-00-03-RO1 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 10 CFR 50 Appendix R Sections III.G.1 and III.G.2. The NRC has accepted the BWROG position and issued an SER dated Dec. 12, 2000.Applicability Applicable Alignment Statement Aligns Comments None Alignment Basis For NMP1, using the Electromatic Relief Valves (ERVs) to remove decay heat to the torus and low pressure Core Spray system is an acceptable method for achieving safe shutdown.The Emergency Condensers provide one shutdown flow path while the ERVs provide an Alternate Shutdown method.Reference Document EIR 51-9133191, NSCA, Section 4.0 NMP1 Safety Evaluation 84-18, ADS Logic Modification NMPI, April 2013 Page B-7 I I NMP1, April 2013 Page B-7 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.1.1.3 Pressurizer Heaters NEI 00-01 Guidance[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.

Applicability Not Applicable Alignment Statement Not Required Alignment Basis NMP1 is a BWR plant.Reference Document Comments NMP1 is a BWR plant.NMPI, April 2013 Page B-8 I I NMP1, April 2013 Page B-8 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.1.1.4 Alternative Shutdown Classification NEI 00-01 Guidance The classification of shutdown capability as alternative/dedicated shutdown is made independent of the selection of systems used for shutdown.

Alternative/dedicated 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 III.G.2 may be supplemented by the use of operator manual actions to the extent allowed by the regulations and the licensing basis of the plant (see Appendix E), repairs (cold shutdown only), exemptions, deviations, GL 86-10 fire hazards analyses or fire protection design change evaluations permitted by GL 86-10, as appropriate.

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

A discussion of time zero for the fire condition, as it relates to operator manual actions and repairs, is contained in Appendix E.Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis The criteria of 10CFR50 Appendix R, Section III.G.3, requires alternate or dedicated shutdown capability for all plant areas where the protection of systems for hot shutdown does not satisfy the requirements of Sections Ill.G.1 and IlI.G.2 of Appendix R. The hot shutdown system to be used for a control room evacuation event is the EC system.NMP1 has two Remote Shutdown Panels installed for the purpose of monitoring the Plant Shutdown in the event of a Control Room evacuation.

Modifications were performed which hardened the EC system from spurious isolations due to the effects of a control complex fire. Upon receiving either a high reactor pressure signal or a low-low reactor water level signal, the ECs will automatically initiate, independent of the control complex, due to the shutdown supervisory control system redundant initiation logic located in the reactor building, although Operator action will initiate the safe shutdown systems prior to its automatic initiation to conserve reactor vessel inventory.

Reference Document EIR 51-9133191, NSCA, Section 2.1.1 NMP1 Safety Evaluation 83-29, Emergency Condensers NMPI, April 2013 Page B-9 I I NMPI, April 2013 Page B-9 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.1.1.5 Operable and Available NEI 00-01 Guidance 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 normal lineups.Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis As stated in the Fire Area Assessments at the onset of the fire, all systems not affected by the fire are considered to be available and capable of functioning as designed.

All safe shutdown systems (including redundant trains) are assumed to be operational and available for post-fire safe shutdown.

Systems are assumed to be operational with no repairs, maintenance, testing, Limiting Conditions for Operation, etc. The unit is assumed to be operating at full power under normal conditions and normal lineups.Reference Document EIR 51-9133191, NSCA, Section 4.0 NMPI, April 2013 Page B-b I I NMP1, April 2013 Page B-10 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.6 No concurrent DBAs No Final Safety Analysis Report accidents or other design basis events (e.g. loss of coolant accident, earthquake), single failures or nonfire induced transients need be considered in conjunction with the fire.Applicability Comments Applicable None Alignment Statement Aligns Alignment Basis No accidents or other design basis events (e.g. loss of coolant accident, earthquake), single failures or non-fire induced transients are considered in conjunction with the fire.The fire does not occur simultaneously or coincident with any other transient or abnormal condition, except for loss of offsite power (LOOP) and those conditions resulting directly from the effects of a fire. No credit is taken for offsite power availability; however, offsite power is considered to be available if the fire effects produce more conservative results.Reference Document EIR 51-9133191, NSCA, Section 9.0 NMPI, April 2013 Page B-lI I I NMP1, April 2013 Page B-11 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.1.1.7 Offsite Power Availability NEI 00-01 Guidance 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 />.Applicability Applicable Comments ONS RAI 3-40 is addressed; No credit is taken for offsite power.Alignment Statement Aligns Alignment Basis For redundant safe shutdown, offsite power is presumed lost. However, offsiteOffsite power is 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 is taken for a fire causing a loss of offsite power. For areas where train separation cannot be achieved and alternative shutdown capability is necessary, shutdown is demonstrated both where offsite power is available and where offsite power is not available (NFPA 805 requires maintaining the fuel in a safe and stable condition, i.e., there is no requirement to achieve and maintain CSD, and therefore no 72-hour coping time requirement). -f9 72 hheur.-Offsite power has not been specifically analyzed.

There are no fire areas where offsite power is credited.Reference Document EIR 51-9133191, NSCA, Section 9.0 ONS RAI 3-40, NRC Request for Additional Information dated July 30, 2010 (ML102110394)

Closure of National Fire Protection Association 805 Transition Program Frequently Asked Question Number 08-0054 (ML110140183)

NMPI, April 2013 Page 6-12 I I NMP1, April 2013 Page B-12 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.8 Safety Related Classification Post-fire safe shutdown systems and components are not required to be safety-related.

Applicability Comments Applicable None Alignment Statement Aligns Alignment Basis Credited safe shutdown systems and components are not always safety related. Most components are safety related due to their emergency functions, but they are not required to be safety-related.

Reference Document EIR 51-9133191, NSCA, Section 4.0 NMPI, April 2013 Page B-13 I I NMP1, April 2013 Page B-13 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.1.1.9 72-hour Coping Period NEI 00-01 Guidance 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.Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis Appendix R requires cold shutdown of the reactor within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for fire events, with or without offsite power available.

The original NMP1 The-Safe Shutdown Analysis identifies the safe shutdown systems and components which are powered by on-site sources, where at least one train can be repaired or made operable within the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. However, NFPA 805 does not require a 72-hour coping period. Rather, the requirement is to maintain the fuel in a safe and stable condition (i.e., there is neither a requirement nor timeframe to reach and maintain cold shutdown).

Refer to Section 4.2.1.2 for a description of safe and stable as applied to NMP1. NMP1 demonstrates the ability to maintain for This is ac..mplished for oach fire area the fuel in a safe and stable condition with one of four designated shutdown paths."mld Id-tains. Offsite power is not credited with providing any power or beneficial effects.ahutdaum m.thods duri~n the 72 hMUMs. thereb mot--no,.

thi ream'ro.--_.

Shutdown Paths:Cold Shu8tdo-wn, Options: A.- Train 11 -Emergency cooling, Shutdo, n "coling, RBCLC, ESW, CRD system (makeup).B. Train 12 -Emergency cooling, Shutdown -coling, RSCLC, ESW, CRD system (makeup).C. Train 11 -ERVs, core spray, containment spray, containment spray raw water.D. Train 12 -ERVs, core spray, containment spray, containment spray raw water.Note: As part of the NMP1 defense-in-depth approach to fire protection, provisions have been made for permanent installation of a feedwater/fire protection water spool piece, which will provide an emergency makeup source from the diesel fire pump for cold shutdown inventory control.S-hutdownim Coo-ling with LOOP (Options A and 2)The primary means for aahieving and maintaining Gold shutdown following a fire coeinc-ld-ent ith a LOGOP is via the shutdown cooling system. The shutdown cooling system, supported by the RBCLC and the SSW systems, rcmovoc decay heat from the reactor. vesscolet the UHIS. Em~ergency AG power (Oncite DOs) is required for this mode of Geld shutdown.Reference Document EIR 51-9133191, NSCA, Section 4.0 and 5.0 HNP RAI 3-7 and RAI 3.8, NRC Request for Additional Information dated August 6, 2009 (ML092170715)

Closure of National Fire Protection Association 805 Transition Program Frequently Asked Question Number 08-0054 (ML1 10140183)lNMP1, April 2013 Page B-14 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.10 Manual Initiation of Systems Manual initiation from the main control room or emergency control stations of systems required to achieve and maintain safe shutdown is acceptable where permitted by current regulations or approved by NRC (See Appendix E); automatic initiation of systems selected for safe shutdown is not required but may be included as an option, if the additional cables and equipment are also included in the analysis.Spurious actuation of automatic systems (Safety Injection, Auxiliary Feedwater, High Pressure Coolant Injection, Reactor Core Isolation Cooling, etc.) due to fire damage, however, should be evaluated.

Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis Manual initiation of safe shutdown related equipment from either the control room, emergency control stations or approved locations other than the primary control stations is an acceptable means for compliance based on the current regulations.

The safe shutdown performance and design requirements for the reactivity control function can be met without automatic scram capability.

The post-fire safe shutdown analysis need only provide the capability to manually scram the reactor. Automatic functions of components have been included for selected systems. Impacts due to spurious actuation of automatic systems are included in the evaluation of the analysis.Reference Document EIR 51-9133191, NSCA, Sections 4.0, 5.1, 5.2, and 5.4 NMP1, April 2013 Page B-15 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.1.1.11 Multi-unit Plant NEI 00-01 Guidance 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.

Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis As stated in the UFSAR, simultaneous fires affecting NMP1 and NMP2 are not anticipated, due to the spatial separation between the two units and the designed separation (fire barrier) at facility interfaces.

NMP1 and NMP2 do not share common facilities for the support of reactor operation or generation of electricity.

However, there is the capability to cross-tie the firewater system between NMP1 and NMP2 via manual ross-tie valves.Reference Document UFSAR, Appendix 10A, Section 2.1.9 NM P1, April 2013 Page B-16 I!NMP1, April 2013 Page B-16 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref NEI 00-01 Guidance 3.1.2 Shutdown Functions The following discussion on each of these shutdown functions provides guidance for selecting the systems and equipment required for hot shutdown.

For additional information on BWR system selection, refer to GE Report GENE-T43-00002-00-01-ROl entitled "Original Safe Shutdown Paths for the BWR." Applicability Comments Applicable None Alignment Statement Not Required Alignment Basis This paragraph provides introductory information and contains no specific requirements.

Discussion is provided in subsequent sub-sections.

Reference Document NMPI, April 2013 Page B-17 I I NMP1, April 2013 Page B-17 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.1.2.1 Reactivity Control NEI 00-01 Guidance[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. Each licensee should have an operator manual action to either vent the instrument air header or to remove RPS power in their post-fire safe shutdown procedures.

The presence of this action precludes the need to perform circuit analysis for the reactivity control function and is an acceptable way to accomplish this function.

If this action is a "time critical" action, the timing must be justified.

[PWR] Makeup/Charging There must be a method for ensuring that adequate shutdown margin is maintained from initial reactor SCRAM to cold shutdown conditions, by controlling Reactor Coolant System temperature and ensuring borated water is utilized for RCS makeup/charging.

Applicability Applicable Comments None Alignment Statement Aligns With Intent Alignment Basis As documented in the current NMP1 Safe Shutdown Analysis, the reactivity control function is capable of achieving and maintaining cold shutdown reactivity conditions.

The safe shutdown performance and design requirements for the reactivity control function can be met without automatic scram capability.

The post-fire safe shutdown analysis provides the capability to manually scram the reactor. Manual reactor scram is accomplished via the scram valves. Plant operators can manually scram the plant from the control room or the remote shutdown panel. The capability to manually scram/trip the reactor is provided in NMP1 Special Operating Procedures N1-SOP-21.1, Fire in Plant, and N1-SOP-21.2, Control Room Evacuation, with reference to the Emergency Operating Procedures for operators to manually vent the instrument air header or to remove RPS power. This action is not considered a "time critical" action because the Mode switch is placed in shutdown and all control rods inserted prior to evacuating the control room. Also, NMP1 complies with the position in BWROG document BWROG-TP-1 1-011 entitled "BWROG Assessment of Generic Multiple Spurious Operations (MSOs) in Post-Fire Safe Shutdown Circuit Analysis for the Operation of BWR Plants" for manual scram; thereby, supporting that this is not a time critical action.Reference Document EIR 51-9133191, NSCA, Sections 4.0 and 5.3 NMP1 Emergency Operating Procedure N1-EOP-2, RPV Control NMP1 Emergency Operating Procedure 3, Failure to Scram NMP1 Emergency Operating Procedure 3.1, Alternate Control Rod Insertion NMP1 Special Operating Procedure, N1-SOP-21.2, Control Room Evacuation, pg. 3 NMP1 Special Operating Procedure, N1-SOP-21.1, Fire in Plant, pg. 2 NMPI, April 2013 Page B-18 I I NMP1, April 2013 Page B-18 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.2 Pressure Control Systems 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)Initial pressure control may be provided by the SRVs mechanically cycling at their setpoints (electrically cycling for EMRVs).Mechanically-actuated SRVs require no electrical analysis to perform their overpressure protection function.

The SRVs may also be 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 (ADS) is not a required function.

Automatic initiation of the ADS may be credited, if available.

If automatic ADS is not available and use of ADS is desired, an alternative means of initiation of ADS separate from the automatic initiation logic for accomplishing the pressure control function should be provided.[PWR] Makeup/Charging RCS 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.

Applicability Comments Applicable None Alignment Statement Aligns Alignment Basis The pressure control function is capable of safely reducing reactor vessel pressure.For the preferred shutdown method, pressure control is achieved through control of the cooldown rate of the emergency condensers.

The Electromatic Relief Valves (ERVs) are maintained closed.In the event the preferred shutdown method is not available, the plant can be shut down by opening three ERVs and discharging steam to the Torus to reduce pressure.

For this secondary method, the ERVs are opened manually to depressurize the vessel to allow injection using low pressure systems. Automatic initiation of the ADS is not a required function.Reference Document EIR 51-9133191, NSCA, Sections 4.0, 5.1, 5.2, and 5.3 NMPI, April 2013 Page 6-19 I I NMP1, April 2013 Page B-19 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.1.2.3 Inventory Control NEI 00-01 Guidance[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.

Spurious actuation of automatic systems, however, should be evaluated (High Pressure Coolant Injection, High Pressure Core Spray, Reactor Core Isolation Cooling, etc.).[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.

Spurious actuation of automatic systems, however, should be evaluated (Safety Injection, High Pressure Injection, Auxiliary Feedwater, Emergency Feedwater, etc.).Applicability Applicable Alignment Statement Aligns Alignment Basis Comments None For the preferred hot shutdown method, reactor vessel make-up is required 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after operation of the Emergency Condensers (EC) ensues. A Control Rod Drive pump is used to provide vessel make-up. The inventory makeup provided by the CRD pump is also used part of Gold shuJtdo, to raise the Reactor water level at such time that and-make-the shutdown cooling system is placed in offec"Ps,.

Restoration of the CRD pump is vi. r.pair a.... to ,ecur po.er SOUrcs, SIc.NMP1 has a defense-in-depth inventory control approach for fire protection by using the firewater to feedwater connection in accordance with NMP1 operating procedures, which provides an emergency makeup source from the diesel fire pump for Gold-shutdown inventory control.As a secondary method, the vessel can be flooded by the Core Spray system after depressurization.

This fulfills the vessel inventory make-up by default.Spurious Actuations are addressed specifically in the Fire Area Assessments.

They are considered to exist from the onset of the fire and for the duration of the shutdown process.Reference Document EIR 51-9133191, NSCA, Sections 5.3 and 8.6 NMP1, April 2013 Page B-20 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.4 Decay Heat Removal [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. (This is not a hot shutdown requirement).

[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 is not a hot shutdown requirement).

This does not restrict the use of other systems.Applicability Comments Applicable None Alignment Statement Aligns Alignment Basis The decay heat removal function is capable of achieving and maintaining safe and stable conditions.hot and cGo-ld shutdoWn.

The EC system operates by natural circulation where steam flows upward to the condenser(s) and returns as condensate to the Reactor Pressure Vessel. Decay heat is removed through the transfer of heat from the reactor coolant to the shell side water of the EC which vents the developed steam to atmosphere.

Operation of either EC loop can sustain Hot Shutdown conditions for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> without the need for makeup from the Condensate Storage Tank or Fire Water System. The decay heat removal process also reduces reactor pressure.

When reactor pressure is reduced to 120 psig and reactor temperature is reduced to 350 degrees F, the plant can be transitioned to Cold Shutdo.w by inifiating shutdown cooling. The EC system can be used to maintain Hot Shutdown conditions for an extended period of time by using the diesel fire pump to refill the Emergency Condensers as needed. The diesel fire pump draws water from Lake Ontario (essentially an infinite source) and need only be refueled periodically to maintain this method of decay heat removal. Note that Emergency Condenser level monitoring is a credited function.In the event the preferred shutdown method is not available, the plant can be shut down by opening three ERVs and discharging steam to the Torus to reduce pressure.

When reactor pressure reaches approximately 365 psig, Core Spray (CS) may be initiated.

CS is a two loop system. Operation of one loop is adequate to achieve shutdown.

Eventually, the reactor vessel floods to the point where the ERVs are passing fluid to the Torus rather than steam, in essence placing the Reactor Coolant System in recirculation through the Torus.During this process, decay heat is removed by operation of the Containment Spray System in conjunction with the Containment Spray Raw Water system. Containment spray and containment spray raw water systems are utilized to remove heat from the torus water and maintain it within the core spray and containment spray pumps' net positive suction head (NPSH) requirements.

This shutdown method will bring the plant directly to cold shutdown.

Fully flooding the Reactor Pressure Vessel negates the need for another system to provide inventory makeup. AC power is required to initiate this shutdown method.Reference Document EIR 51-9133191, NSCA, Sections 4.0 and 5.1 NMPI, April 2013 Page B-21 I!NMP1, April 2013 Page B-21 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.1.2.5 Process Monitoring NEI 00-01 Guidance 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 (10 CFR 50 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 1I.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/dedicated shutdown (III.G.3).

The use of this same list for Ill.G.2 redundant Post-Fire Safe Shutdown is acceptable, but the analyst needs to review the specific license basis for the plant under evaluation.

In general, process monitoring instruments similar to those listed below are needed to successfully use existing operating procedures (including Abnormal 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.Comments None Applicability Applicable Alignment Statement Aligns Alignment Basis The process monitoring function is to be provided for all safe shutdown paths. NEI 00-01 refers to NRC IN 84-09, Attachment 1,Section IX, as providing 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 III.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 general, process monitoring instruments similar to those listed below are needed to successfully use existing operating procedures related to post-fire shutdown (including Abnormal Operating Procedures).

I NMP1, April 2013 Page B-22 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review-Reactor coolant level and pressure-Suppression pool level and temperature-Level indication for tanks needed for safe shutdown-Diagnostic instrumentation for safe shutdown systems The Reactor Protection System is utilized to satisfy the process monitoring objectives throughout hot and cold shutdown.

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.The following process monitoring functions are provided to support post-fire shutdown: Primary and Secondary Methods-Reactor coolant level-Reactor coolant pressure-Reactor coolant temperature-Emergency Diesel Generator parameters Primary Method-Emergency Condenser level Secondary Method-Torus level-Torus temperature-Drywell pressure-Drywell temperature-Containment Spray water temperature-Containment Spray pump discharge pressure Reference Document EIR 51-9133191, NSCA, Sections 4.0 and 5.3 NMPI, April 2013 Page B-23 I!NMP1, April 2013 Page B-23 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.6 Support Systems Blank Heading -No Specific Guidance Applicability Comments Applicable None Alignment Statement Not Required Alignment Basis Generic Heading: Alignment discussed in subsequent sub-sections.

Reference Document NMPI, April 2013 Page B-24 I NMPI, April 2013 Page B-24 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.6.1 Electrical 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 any beneficial effects of 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 125VDC distribution panels may also supply power to the 120VAC distribution panels via static inverters.

These distribution panels may 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 125VDC distribution system can be powered from sources feed from the diesels through the battery chargers.[BWR] Certain plants are also designed with a 250VDC Distribution System that supplies power to Reactor Core Isolation Cooling and/or High Pressure Coolant Injection equipment.

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).Applicability Comments Applicable None Alignment Statement Aligns Alignment Basis No credit has been taken for the loss of offsite power, however in the event that offsite power is lost, NMP1 has two Emergency Diesel Generators (EDGs). The EDGs will supply the required medium and low voltage safe shutdown loads with AC power. The EDGs are designed to start automatically on loss of offsite power to re-energize emergency busses 102 and 103.The 125V DC distribution panels supply power to the 120V AC distribution panels via static inverters.

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

The 125V DC distribution system supplies control power to various 125V DC control panels including switchgear breaker controls.

Vital AC power can be provided via RPS uninterruptible power supplies (UPS) 162A, 162B, 172A and 172B.This 125V DC distribution system is credited to support post fire safe shutdown.

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

Once the EDGs are operational, the 125V DC distribution system can be powered from the diesels through the battery chargers.Reference Document EIR 51-9133191, NSCA, Sections 4.0 and 5.3!NMP1, April 2013 Page B-25 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.6.2 Cooling Systems Various cooling water systems may be required to support safe shutdown system operation, based on plant-specific considerations.

Typical uses include:* RHR/SDC/DH Heat Exchanger cooling water* Safe shutdown pump cooling (seal coolers, oil coolers)* Diesel generator cooling Applicability Comments Applicable None Alignment Statement Aligns Alignment Basis Various cooling water systems are required to support safe shutdown system operation:-The Reactor Building Closed Loop Cooling and Emergency Service Water Systems provide cooling to the control room ventilation system and Shutdown Cooling Heat Exchangers-The EDGs are cooled by the EDG Raw Water pumps-The Containment Spray and Containment Spray Raw Water Systems cool the torus for the secondary cooldown method-The Chilled Water System is provided for control room ventilation Reference Document EIR 51-9133191, NSCA, Section 5.3 NMPI, April 2013 Page B-26 I I NMP1, April 2013 Page B-26 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.6.3 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, however, are not required to support post-fire safe shutdown in all cases. The need for HVAC system operation is based on plant specific configurations and plant specific heat loads. Typical potential 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 could be required or useful in supporting post-fire safe shutdown.

Transient temperature response analyses are often utilized to demonstrate that specific HVAC systems would or would not be required.

If HVAC systems are credited, the potential for adverse fire effects to the HVAC system must also be considered, including:

  • Dampers closing due to direct fire exposure or due to hot gases flowing through ventilation ducts from the fire area to an area not directly affected by the fire. Where provided, smoke dampers should consider similar effects from smoke.* Recirculation or migration of toxic conditions (e.g., smoke from the fire, suppressants such as Carbon Dioxide).In certain situations, adequate time exists to open doors to provide adequate cooling to allow continued equipment operation.

Therefore, the list of required safe shutdown components as it relates to HVAC Systems may be determined based on transient temperature analysis.

Should this analysis demonstrate that adequate time exists to open doors to provide the necessary cooling, this is an acceptable approach to achieving HVAC Cooling. The temperature analysis must demonstrate the adequacy of the cooling effect from opening the door within the specified time. Only those components whose operation is required to provide HVAC Cooling for required safe shutdown components in a time frame that cannot be justified for operator manual actions are considered themselves to be required safe shutdown components.

This latter set of HVAC Cooling Components are required to meet the criteria for required safe shutdown components with regard to the available mitigating tools.Applicability Comments Applicable None Alignment Statement Aligns Alignment Basis HVAC Systems required to support post-fire shutdown are:-Main control room ventilation (except for control room evacuation)-EDG rooms (fans and roll-up doors)Paths developed for the control room ventilation system do not necessarily match the paths for process systems. Paths for the ventilation system are as follows:-Path A -Recirculation flow path-Path B -Smoke purge flow path I NMPI, April 2013 Page B-27 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review-Path C -Train 11 positive pressure flow path.-Path D -Train 12 positive pressure flow path Given the failure modes of dampers and EDG power requirement, Paths C & D are generally credited for post-fire shutdown.

Any ventilation path can be supported by Train 11 or Train 12 Chilled Water.Reference Document EIR 51-9133191, NSCA, Section 5.3 NMPI, April 2013 Page B-28 I!NMPI, April 2013 Page B-28 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.1.3 Methodology For Shutdown System Collection NEI 00-01 Guidance 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: Applicability Applicable Comments None Alignment Statement Not Required Alignment Basis This paragraph provides introductory information and contains no specific requirements.

Discussion is provided in subsequent sub-sections.

Reference Document NMPI, April 2013 Page B-29 I I NMPI, April 2013 Page B-29 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.1.3.1 Identify Safe Shutdown Functions NEI 00-01 Guidance 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)[BWR] GE Report GE-NE-T43-00002-00-01-R02 entitled "Original Shutdown Paths for the BWR" Applicability Applicable Alignment Statement Aligns Comments None Alignment Basis Safe shutdown systems and functions required to satisfy the safe shutdown performance goals were developed and identified from available plant documentation.

This documentation includes but is not limited to electrical one line diagrams, schematics, piping and instrumentation diagrams (P&lDs), operating procedures, UFSAR, Fire Hazards Analysis, and the systems descriptions.

Reference Document EIR 51-9133191, NSCA, Section 5.4 NM P1, April 2013 Page B-30 I I NMP1, April 2013 Page B-30 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.1.3.2 Identify Combinations of Systems That Satisfy Each Safe Shutdown Function Applicability Applicable Alignment Statement Aligns NEI 00-01 Guidance 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 other equipment whose mal-operation or spurious operation could impact the required safe shutdown function.

The components in this latter set are classified as either required for hot shutdown or as important to SSD as explained in Appendix H.Comments Alignment Basis Safe shutdown systems, components, and cables were selected using the criteria and assumptions of NEI 00-01, Sections 3.1.1 and 3.1.2 to satisfy each safe shutdown function performance goal, including process monitoring and support systems. NEI 00-01 was used as the guidance in developing the Safe Shutdown Equipment List and the Safe Shutdown success paths.The following combination of systems are capable of achieving safe shutdown functions.

1 ) Reactivity Control -Control Rod Drive System The safe shutdown performance and design requirements for the reactivity control function can be met without automatic scram capability.

The post-fire safe shutdown analysis need only provide the capability to manually scram the reactor.2) Pressure Control Systems -ERVs The ERVs 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 ADS is not a required function.3) Inventory Control Systems selected for the inventory control function are capable of supplying sufficient reactor coolant to maintain the fuel in a safe and stable condition.a.hiev...

and maintain hot shuewri.-

Manual initiation of these systems is acceptable.

Automatic initiation functions are not required.4) Decay Heat Removal Systems selected for the decay heat removal function(s) are 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 tows.Removing sufficient decay heat from the reactor to maintain the fuel in a safe and stable condition.aa-.hia-' sh'tdo'wn.

!NMP1, April 2013 Page B-31 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 5) Process Monitoring The process monitoring function is to be provided for all safe shutdown paths. NEI 00-01 refers to NRC IN 84-09, Attachment 1,Section IX, as providing 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 III.L.I. Such instrumentation must be demonstrated to remain unaffected by the fire. The IN 84-09 list of process monitoring functions is applied to alternative shutdown (III.G.3).

In general, process monitoring instruments similar to those listed below are needed to successfully use existing operating procedures related to post-fire shutdown (including Abnormal Operating Procedures).

-Reactor coolant level and pressure-Torus level and temperature

-Level indication for tanks needed for safe shutdown-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.Note -For NMP1, Emergency Condenser level is a required process monitoring function.6) Support Systems A. Electrical Systems AC Distribution System Power for safe shutdown equipment is supplied from either offsite power sources or the emergency diesel generator.

No credit is taken for a fire causing a loss of offsite power.DC Distribution System The 125V DC distribution system supplies control power to various 125V DC control panels including switchgear breaker controls.

The 125V DC distribution panels also supply power to the 120V AC distribution panels via static inverters.

These distribution panels supply power for instrumentation necessary to complete 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 125V DC distribution systems can be powered from the diesels through the battery chargers.B. Cooling Systems Various cooling water systems are required to support safe shutdown system operation, based on plant-specific considerations.

Cooling system uses include:-SDC Heat Exchanger cooling water-Safe shutdown pump cooling (seal coolers, oil coolers)-Diesel generator cooling-HVAC system cooling water C. HVAC Systems HVAC Systems are 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).NMPI, April 2013 Page B-32 I I NMP1, April 2013 Page B-32 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review HVAC uses include:-Main control room, cable spreading room, relay room-ECCS pump compartments

-Diesel generator rooms-Switchgear Rooms Paths developed for the control room ventilation system do not necessarily match the paths for process systems. Paths for the ventilation system are as follows:-Path A -Recirculation flow path-Path B -Smoke purge flow path-Path C -Train 11 positive pressure flow path.-Path D -Train 12 positive pressure flow path Given the failure modes of dampers and EDG power requirement, Paths C & D are generally credited for post-fire shutdown.

Any ventilation path can be supported by Train 11 or Train 12 Chilled Water.Reference Document EIR 51-9133191, NSCA, Sections 4.0, 5.0, and 5.4 NMPI, April 2013 Page B-33 I I NMP1, April 2013 Page B-33 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.1.3.3 Define Combination of Systems for Each Safe Shutdown Path NEI 00-01 Guidance 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.Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis Safe shutdown systems, components, and cables were selected using the criteria and assumptions of NEI 00-01, Sections 3.1.1 and 3.1.2 to satisfy each safe shutdown function performance goal, including process monitoring and support systems for each path. NEI 00-01 was used as the guidance in developing the Safe Shutdown Equipment List and the Safe Shutdown success paths. The shutdown paths and equipment selection for the shutdown performance goals are identified in the NSCA.Reference Document EIR 51-9133191, NSCA, Sections 4.0 and 5.0 NMPI, April 2013 Page 6-34 I!NMP1, April 2013 Page B-34 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.1.3.4 Assign Shutdown Paths to Each Combination of Systems NEI 00-01 Guidance 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 I to this document for an example of a table illustrating how to document the various combinations of systems for selected shutdown paths.Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis Safe shutdown systems, components, and cables were selected using the criteria and assumptions of NEI 00-01, Sections 3.1.1 and 3.1.2 to satisfy each safe shutdown function performance goal, including process monitoring and support systems for each path. NEI 00-01 was used as the guidance in developing the Safe Shutdown Equipment List and the Safe Shutdown success paths. Safe Shutdown Paths were designed based on the combination of systems in the respective fire area.The major systems for safe shutdown success paths are as follows: SUCCESS PATH "A" Hot shutdown is achieved via the use of ECs 111 & 112 for Decay Heat Removal. The DFP is used as an inventory source for the EC system. Cold ANhutdwn is aseempliched Aia the use of Tra-in 11 SDC System components supported by the T-Fain I1 RBCLC and ESW Systems. The Train 11 CRD pump or the DFP is used for inventory control/RPV makeup. Path A is generally supported by the Train 11 AC and DC power systems. Some components in Path A systems are powered by the Train 12 AC and/or DC power system while other components may require local operator action due to a potential loss of instrument air.SUCCESS PATH "B" Hot shutdown is achieved via the use of ECs 121 & 122 for Decay Heat Removal. The DFP is used as an inventory source for the EC system ...ld. h..dWn.. is acmp.sh.d vA th, use of Tahin 12 SDC System campnotSM.

Supported by the TrFain 12 RBCLC and ESW Systems. The Train 12 CRD pump or the DFP is used for inventory control/RPV makeup. Path B is generally supported by the Train 12 AC and DC power systems. Some components in Path B systems are powered by the Train 11 AC and/or DC power system while other components may require local operator action due to a potential loss of instrument air.SUCCESS PATH "C" Rath hot and cold shu tdo', aroHot shutdown is achieved via use of the Train 11 ERVs (PSV-01-102A, PSV-01-102B, PSV-01-102E) to depressurize the reactor. When pressure drops to the appropriate level, the CS System is used to flood the vessel and place the RCS in a recirculation mode to the Torus. Torus cooling is provided by the CTS system supported by the CTSRW System. These systems can be are-maintained in service to directly achieve cold shutdown conditions.

Path C is generally supported by the Train 11 AC and DC power systems. Some components in Path C systems are powered by the Train 12 AC and/or DC power system.NMPI, April 2013 Page B-35 I I NMP1, April 2013 Page B-35 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review SUCCESS PATH "D" Hot shutdown is -ot- hot chu-tdo4-.

and "old h-chutdo.a, are achieved via use of the Train 12 ERVs (PSV-01-102C, PSV-01-102D, PSV-01-102F) to depressurze the reactor. When I pressure drops to the appropriate level, the CS System is used to flood the vessel and place the RCS in a recirculation mode to the Torus. Torus cooling is provided by the CTS System supported by the CTSRW System. These systems can be are-maintained in service to directly achieve cold shutdown conditions.

Path D is generally supported by the Train I 12 AC and DC power systems. Some components in Path D systems are powered by the Train 11 AC and/or DC power system.Path A and Path B are the preferred shutdown paths to maintain the fuel in a safe and stable condition for both hot sh-Atdo.-

and "ald r-hutdoA.-

as these paths present the least I thermal hydraulic impact to the plant.It is possible that for some fire areas, one Path may be employed for hot shutdown and another for cold shutdown.

This depends upon the impacts to power supplies and other components in any particular fire area, the controls design of credited components and the use of damage repair procedures.

Consequently, it is possible that Path B may be credited for hot shutdown and Path D for cold shutdown, or any other combination.

However, if Path C or D is credited for hot shutdown, that path would also be credited for cold shutdown due to the dynamics of the shutdown process.Reference Document EIR 51-9133191, NSCA, Section 5.3 Closure of National Fire Protection Association 805 Transition Program Frequently Asked Question Number 08-0054 (ML110140183)

NMPI, April 2013 Page B-36 IlNMP1, April 2013 Page B-36 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.2 Safe Shutdown Equipment Selection NEI 00-01 Guidance 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 functions.

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 mal-operation of the safe shutdown systems. For each path it will be important to understand which components are classified as required safe shutdown components and which are classified as important to safe shutdown components.

When evaluating the fire-induced impact to each affected cable/component in each fire area, this classification dictates the tools available for mitigation the affects.Applicability Applicable Comments None Alignment Statement Not Required Alignment Basis This paragraph provides introductory information and contains no specific requirements.

Discussion is provided in subsequent sub-sections.

Reference Document NMPI, April 2013 Page B-37 I I NMP1, April 2013 Page B-37 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref NEI 00-01 Guidance 3.2.1 Criteria/Assumptions Consider the following criteria and assumptions when identifying equipment necessary to perform the required safe shutdown functions:

Applicability Comments Applicable None Alignment Statement Not Required Alignment Basis This paragraph provides introductory information and contains no specific requirements.

Discussion is provided in subsequent sub-sections.

Reference Document NMPI, April 2013 Page B-38 I!NMP1, April 2013 Page B-38 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.1 Safe Shutdown Equipment Safe shutdown equipment can be divided into two categories.

Equipment may be categorized as (1) primary components or (2) secondary Categories 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.Applicability Comments Applicable None Alignment Statement Aligns Alignment Basis The Safe Shutdown Equipment List (SSEL) is a list of analyzed components that are utilized in the post-fire safe shutdown analysis to ensure that one success path (structures, systems, and components) necessary to achieve safe shutdown is free of fire damage without crediting plant or system repair.The current NMP1 SSEL and Safe Shutdown Analysis was reviewed against the criteria outlined in NEI 00-01 (Sections 3.1 and 3.2) for safe shutdown systems and equipment selection.

This review addressed potential fire induced circuit failure issues, either within or beyond the plant's existing licensing basis. Additional equipment has been included to address any multiple spurious operation concerns.The SSEL is divided into primary and secondary components.

Primary and secondary components were defined as being consistent with NEI 00-01 guidance.

Equipment identified as primary components is included in the SSEL. Equipment identified as secondary components is included in the SSEL Database with the primary component(s) that would be affected by fire damage to the secondary component.

By doing this, the SSEL is kept to a manageable size and the equipment included in the SSEL can be readily related to required post-fire safe shutdown systems and functions.

Secondary components were generally combined with primary components except where groups of secondary components were defined as "pseudo-components." A "pseudo-component" is an artificial association of equipment and cables that perform a common function as a single entity for analysis purposes.

The "pseudo-component" is assigned for analysis purposes only and is not an actual plant hardware designation.

The concept of a "pseudo-component" was developed to account for those cables which constitute a circuit common to several components.

The use of "pseudo-components" precludes the need to repeat cable selection and circuit analysis of these cables for each primary component.

This generic name is interlocked with the affected primary components for analysis purposes and it inherits the attributes (path, system, train, etc.) of the components that it may affect. The nomenclature of the "pseudo-component" is similar to other equipment as defined in the plant equipment database.NMPI, April 2013 Page B-39 I I NMPI, April 2013 Page B-39 I Constellation Energy Nuclear Group Reference Document EIR 51-9133191, NSCA, Section 5.4 EIR 51-9177678-000, Definitions Section Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NMPI, April 2013 Page B-40 I NMP1, April 2013 Page B-40 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.2.1.2 Manual Valves and Piping NEI 00-01 Guidance 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.

For example, post-fire coefficients of friction for rising stem valves cannot be readily determined.

Handwheel sizes and rim pulls are based on well lubricated stems. Any post-fire operation of a rising stem valve should be well justified using an engineering evaluation.

Applicability Applicable Comments Instrument tubing failure damage due to a fire is addressed in NSCA ( EIR 51-9133191, Section 8.4).Alignment Statement Aligns With Intent Alignment Basis The NMP1 fire area assessments assume that an exposure fire does not adversely affect the ability of manual valves and piping to perform their pressure boundary or safe shutdown function.

Fire damage to valves has been 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.

Post-fire operation of manual valves within the affected fire area has been evaluated in the Fire Area Analysis.Instrument sensing lines were reviewed for susceptibility to physical fire damage that may cause a loss of inventory.

Sensing lines for SSEL components are constructed of either stainless steel or carbon steel. Consequently, they are not susceptible to physical damage as the result of a postulated fire.Reference Document EIR 51-9133191, NSCA, Sections 5.4 and 8.4 HNP RAI 3-15, NRC Request for Additional Information dated August 6, 2009 (ML092170715)

NMPI, April 2013 Page 6-41 I!NMP1, April 2013 Page B-41 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.3 Valves in Normal Position Assume that all components, including manual valves, are in their normal position as shown on P&IDs or in the plant operating procedures, that there are no LCOs in effect, that the Unit is operating at 100% power and that no equipment has been taken out of service for maintenance.

Applicability Comments Applicable None Alignment Statement Aligns Alignment Basis Manual valves are assumed to be in their normal operating position as shown on P&IDs or as identified in the plant operating procedures.

Comments None Reference Document EIR 51-9133191, NSCA, Sections 5.4, 8.1, and 9.0 NMPI, April 2013 Page B-42 I I NMP1, April 2013 Page B-42 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.2.1.4 Check Valves NEI 00-01 Guidance 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.

Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis Check valves are assumed to close in the direction of potential flow diversion and seat properly with sufficient leak tightness to prevent flow diversion or inventory loss. Check valves do not adversely affect flow rate capability of the safe shutdown systems.Comments None Reference Document EIR 51-9133191, NSCA, Section 5.4 NMPI, April 2013 Page B-43 I I NMP1, April 2013 Page B-43 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.2.1.5 Instrument Failure NEI 00-01 Guidance 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.Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis Instruments are assumed to fail in the most undesirable worst state, whether upscale, downscale, or mid-scale.

An instrument providing a control function is assumed to provide an undesired signal to the control circuit.Comments None Reference Document EIR 51-9133191, NSCA, Sections 5.4 and 8.1 NMPI, April 2013 Page B-~ I I NMP1, April 2013 Page B-44 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref NEI 00-01 Guidance 3.2.2 Methodology For Equipment Refer to Figure 3-3 for a flowchart illustrating the various steps involved in selecting safe shutdown equipment.

Use the following Selection methodology to select the safe shutdown equipment for a post-fire safe shutdown analysis: Applicability Comments Applicable None Alignment Statement Not Required Alignment Basis This paragraph provides introductory information and contains no specific requirements.

Discussion is provided in subsequent sub-sections.

Reference Document NMPI, April 2013 Page B-45 I I NMP1, April 2013 Page B-45 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.2.2.1 Identify the System Flow Path for Each Shutdown Path NEI 00-01 Guidance 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 When developing the SSEL, 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.

Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis The Safe Shutdown Equipment List (SSEL) is a list of analyzed components that are utilized in the post-fire safe shutdown analysis to ensure that one success path (structures, systems, and components) necessary to maintain the fuel in a safe and stable condition.'eac-hi cAfe ch.ut-doa:.Afr-e oIs f fire damag. eith"ut Gediting plnt OFr repair. The SSEL is divided into primary and secondary components.

Primary and secondary components were defined as being consistent with NEI 00-01 section 3.2.1.1 guidance.

Secondary components were generally combined with primary components.

Combinations of components and systems with the capability to satisfy the required NFPA 805 performance goals safo funtone were designated as safe shutdown flow paths. P&IDs and Electrical drawings were marked up and annotated to highlight the selected primary safe shutdown equipment and flow paths for each system in support of each shutdown path. The specific group of equipment supporting each system was populated into the safe shutdown database.Reference Document EIR 51-9133191, NSCA, Sections 4.0 and 6.0 I NMPI, April 2013 Page B-46 I!NMP1, April 2013 Page B-46 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.2.2.2 Identify the Equipment in Each Safe Shutdown System Flow Path NEI 00-01 Guidance Review the applicable documentation (e.g. P&IDs, electrical drawings, instrument loop 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.

Additionally, refer to Section 4 for the Resolution Methodology for determining the Plant Specific List of MSOs requiring evaluation.

Criteria for making the determination as to which of these components are to be classified as required for hot shutdown or as important to SSD is contained in Appendix H. If additional systems are identified which are necessary for the operation of the safe shutdown system under review, include these as required for hot shutdown systems. Designate these new systems with the same safe shutdown path as the primary safe shutdown system under review (Refer to Figure 3-1).Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis The safe shutdown flow paths identify the primary components that are required to meet the safe shutdown performance goals. The safe shutdown components were compiled based on each system's performance and safe shutdown function.

These components establish the primary safe shutdown flowpath for system operation.

Also included in the safe shutdown flow paths are those components whose spurious operation could impact safe shutdown system operability.

Systems identified as necessary for the operation of the safe shutdown system under review are included in the safe shutdown equipment list and designated with the same shutdown path as the primary safe shutdown system. These components may involve branch flow paths that must be isolated and remain isolated to assure that flow will not be diverted from the primary flow path. The list of primary components may also include selected mechanical components required to support safe shutdown.The criteria used in evaluating spurious actuation of components are those identified in NEI 00-01, Section 4, Identification and Treatment of Multiple Spurious Operations (MSO). The Nuclear Safety Capability Fire Area Assessments includes the potential impact of multiple spurious component actuations per the guidance of NEI 00-01. MSO component combinations, as documented in the Technical Report on Identification

& Classification of the NMP1 MSO Scenarios Using an Expert Panel -Review of New Generic Scenarios, were addressed in EIR 51-9133191 and included in the fire area assessments.

Reference Document EIR 51-9133191, NSCA, Sections 4.0, 5.0, 6.0, 8.6.1, and 8.6.3 Technical Report on Identification

& Classification of the NMP-1 MSO Scenarios Using an Expert Panel -Review of New Generic Scenarios NMPI, April 2013 Page B-47 I I NMPI, April 2013 Page B-47 i Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.2.2.3 Assign Safe Shutdown Flow Paths NEI 00-01 Guidance 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. Criteria for making the determination as to which of these components are to be classified as required for hot shutdown or as important to SSD is contained in Appendix H.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 post-fire safe shutdown and it documents various equipment-related attributes used in the analysis.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.Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis The Safe Shutdown Equipment List (SSEL) includes equipment for each system supporting the flow paths needed to achieve safe shutdown.

This equipment, identified from the highlighted P&IDs, includes both the normal and diversion flow paths required to meet the system performance goals and safe shutdown functions.

These components also include valves or equipment that could impact safe shutdown by spuriously operating or whose failure would threaten the capability to achieve safe shutdown.

The components were populated into the SSEL database and assigned to safe shutdown system success paths.During the cable selection process additional support components such as electrical distribution equipment were added to the SSD paths and populated into the database.

The database reports produce tables listing equipment and related information which is very similar to the table provided in Attachment 3, of NEI 00-01. This group of components and the various equipment related attributes makes up the SSEL.Instrument sensing lines for level, pressure, flow, etc. that are exposed to a fire are considered to have the potential of causing erratic or unreliable indication.

The instrument tubing lines were traced and their routes correlated to fire areas. Cable identifications were given to the sensing lines and were subjected to the same compliance issues and analytical techniques as safe shutdown cables and similarly analyzed for separation in the fire area assessments.

Reference Document EIR 51-9133191, NSCA, Sections 4.0, 5.3, 6.0, and 8.4 NMPI, April 2013 Page B-48 I I NMP1, April 2013 Page B-48 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.2.2.4 Identify Equipment Information Required for the Safe Shutdown Analysis NEI 00-01 Guidance 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 concern. Criteria for making the determination as to which of these components are to be classified as required for hot shutdown or as important to SSD is contained in Appendix H.Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis The systems and components of the Safe Shutdown Equipment List (SSEL) were selected to meet the NFPA 805 performance goals to ensure the fuel remains in a safe and stable condition.ashieve post fire chutd~m' a- spo.,fied I 1.CFR5O, Appendix R. Additional secondary components which are modeled as a result of the primary component selections were also populated into the database.This additional equipment related information necessary to perform the Fire Area Assessments was collected and included in the SSEL. The SSEL contains the following information; system, train, component, component description, path, Hi/Lo pressure interface determination, normal position, hot shutdown position, cold shutdown position, failed electrical position, failed air position, Fire Area, and Fire Zone.The SSEL database contains equipment and related information similar to the information identified in Attachment 3 of NEI 00-01. The SSEL contains the primary components which are required for hot shutdown.

The secondary components are typically items found within the circuitry for a primary component and provide a supporting role. Components that are important to safe shutdown are all components not classified as required for hot shutdown.Reference Document EIR 51-9133191, NSCA, Sections 4.0 and 6.0 NMPI, April 2013 Page B-49 I I NMP1, April 2013 Page B-49 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection NEI 00-01 Ref 3.2.2.5 Identify Dependencies Between Equipment, Supporting Equipment, Safe Shutdown Systems and Safe Shutdown Paths.NEI 00-01 Guidance In the process of defining equipment and cables for safe shutdown, identify additional 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.

Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis As part of the process of preparing the Safe Shutdown Equipment List (SSEL) and defining equipment and cables for safe shutdown, additional equipment and cables that support the SSEL components were identified (such as interlocked components, normal and alternate electrical power supplies, cascading power supplies).

The process included development of a cascading interlock analysis.This information was populated into a relational type database necessary to analyze for post-fire safe shutdown.Reference Document EIR 51-9133191, NSCA, Sections 4.0, 5.0, and 6.0 NMPI, April 2013 Page B-50 IlNMP1, April 2013 Page B-50 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 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 mal-operation 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.

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.(b) Common Enclosure Circuits.

Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified.

The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref 3.3 Safe Shutdown Cable Selection And Location NEI 00-01 Guidance This section provides industry guidance on one acceptable approach 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 mal-operation 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. The classification of a cable as either an important to SSD circuit cable or a required safe shutdown cable is also derived from the classification applied to the component that it supports.

This classification can vary from one fire area to another depending on the approach used to accomplish post-fire safe shutdown in the area.Refer to Appendix H for the criteria to be used for classifying required and important to SSD components.

Applicability Applicable Comments None Alignment Statement Not Required Alignment Basis This paragraph provides introductory information and contains no specific requirements.

Discussion is provided in subsequent sub-sections.

Reference Document NMPI, April 2013 Page B-51 I NMP1, April 2013 Page B-51 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref 3.3.1 Criteria/Assumptions NEI 00-01 Guidance 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:

Applicability Applicable Comments None Alignment Statement Not Required Alignment Basis This paragraph provides introductory information and contains no specific requirements.

Discussion is provided in subsequent sub-sections.

Reference Document NMPI, April 2013 Page B-52 I I NMP1, April 2013 Page B-52 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref 3.3.1.1.1 Cable Failures NEI 00-01 Guidance 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.

Review 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.Applicability Applicable Alignment Statement Aligns Comments None Alignment Basis The cables necessary to operate and/or maintain the status of each Safe Shutdown component were evaluated by a detailed review of the drawings.

Cables that could impact safe shutdown equipment were identified using the component's control schematic, instrument loop, wiring diagram, if available, or the component's electrical elementary wiring diagrams, one-line drawings, or other available wiring diagrams.

These drawings were used as a guide to perform a point to point review of the associated connection diagrams.

Cables associated with power, control, instrumentation, indication, interlock and any other cable that could impact the component were considered.

Fault analysis during cable identification led to the cable fault codes P, L, 0, C, and I as defined in EIR 51-9133191.

This made the final compliance analysis bounding.

Further analysis determined the effects of a fire induced hot short, open circuit and short to ground during the fire area compliance assessment task. Additional schematic diagrams were reviewed for secondary or interlocked circuits, as necessary, which could impact the operation of components required for safe shutdown.Reference Document EIR 51-9133191, NSCA, Sections 2.6, 7.0, and 8.0 CNG-FES-017, NFPA 805 Safe Shutdown Equipment Cable Selection NMPI, April 2013 Page B-53 I I NMP1, April 2013 Page B-53 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref 3.3.1.1.2 Cable Failures Affecting Multiple Safe Shutdown Equipment NEI 00-01 Guidance In cases where the failure (including spurious operations) of a single cable could impact more than one piece of safe shutdown equipment, associate the cable with each piece of safe shutdown equipment.

Applicability Applicable Alignment Statement Comments None Aligns With Intent Alignment Basis For cases where single cables have the potential to impact multiple components, the cable would be listed against each component.

For control logic circuits, where multiple components receive signals from common control logic, the control logic was analyzed as a primary component and given a pseudo-component identification.

A pseudo-component is an artificial association of equipment and cables that performs a common function that is combined into a single entity for analysis purposes only and is not an actual plant hardware designation.

The pseudo-component was interlocked to the other associated primary components so that the effect of the control logic could be evaluated on an individual component level.This methodology was used for similar circuit scenarios such as common power supplies.

Whereas this approach does not assign the cable to each individual component, the effect on each component due to fire damage is analyzed.

This method serves to reduce the duplication of cable data when the same cables are assigned to multiple components.

Pseudo-components and other primary components, whose associated cabling can affect another primary component based on interposing contacts, were identified on the Cable Selection Worksheet of the affected component as an interlocked primary component.

Reference Document EIR 51-9133191, NSCA, Section 6.0 EIR 51-9177678-0OW, Definitions Section NMP1, April 2013 Page B-54 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref 3.3.1.1.2.1 Electrical Devices NEI 00-01 Guidance Electrical devices such as relays, switches and signal resistor units are considered to be acceptable isolation devices. In the case of instrument loops and electrical metering circuits, 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.

Refer to Section 3.5 for the types of faults that should be considered when evaluating the acceptability of the isolation device being credited.Applicability Applicable Alignment Statement Aligns Comments None Alignment Basis Cables were identified and selected using the component's control schematic, electrical elementary diagrams, one-lines, and/or instrument loop diagrams.

These drawings were used as a guide while performing a point to point review of the associated connection diagrams.Electrical isolation devices prevent malfunctions in one section of a circuit from causing unacceptable effects in other portions of the circuit or other circuits (e.g., open contacts, fuses, switches, instrument isolation modules).

Devices credited as providing electrical isolation were identified in the circuit analysis for the affected component.

Fault analysis during cable identification led to the P, L, 0, C, and I fault codes. All circuits/cables that are electrically connected to the circuit under the analysis are identified up to a credited isolation device including the instrument loops.Reference Document EIR 51-9133191, NSCA, Sections 2.6 and 8.3 NMPI, April 2013 Page B-55 I I NMP1, April 2013 Page B-55 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref 3.3.1.1.3 Screening Out Cables With No Impact NEI 00-01 Guidance 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.

To be properly screened out, however, the circuits associated with these devices 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. Refer to Section 3.5 for the types of faults that should be considered when evaluating the acceptability of the isolation device being credited.Applicability Applicable Alignment Statement Aligns Comments None Alignment Basis Cables necessary to operate and/or maintain the status of each safe shutdown component were identified and evaluated by a detailed review of the component's control schematic, elementary diagrams, one-lines, instrument loop diagrams or other available wiring diagrams.

These drawings were used as a guide while performing a point to point review of the associated connection diagrams.

Cables associated with outputs from auxiliary contacts to computer points, annunciators or motor heaters were excluded from the cable selection when it was concluded that the cable failure would not impact the primary component or performance of the circuit.Panel wires that are completely contained within a panel were not explicitly listed as SSD cables. These wires are inherently included in the analysis in the same manner as secondary components.

Reference Document EIR 51-9133191.

NSCA, Sections 7.0 and 8.3 NMPI, April 2013 Page B-56 I I NMP1, April 2013 Page B-56 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref 3.3.1.1.4 Power Supply to Safe Shutdown Equipment NEI 00-01 Guidance 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. The set of cables described above are classified as required safe shutdown cables.Evaluate the power cables for breaker coordination concerns.

The non-safe shutdown cables off of the safe shutdown buses are classified as required for hot shutdown or as important to SSD based on the criteria contained in Appendix H.Applicability Applicable Alignment Statement Aligns Comments None Alignment Basis The power cables were selected using the component's control schematic or electrical elementary diagrams, one-lines, or instrument loop diagrams or other available wiring diagrams.During the cable selection process, the supporting power sources and interlocks for each primary component were identified.

The cascading power supplies (pseudo-components created for power supply interlocks) and the cascading interlocks all serve to identify required power supplies to ensure safe shutdown components are supplied with electrical power.The relationship between the power source and their load components was documented and their dependency was considered during the Fire Area Assessment phase by reviewing the power source load list report from the database.Breaker coordination calculation 32-9151404-000, Nine Mile Point Unit 1 -NFPA 805 Coordination Study, reviewed the existing and any new electrical circuits that could impact safe shutdown.

This calculation identified fire protection program breaker coordination issues concerning proper coordination between the supply breaker/fuse and the load breaker/fuses for power sources required for hot shutdown.

This document meets the nuclear safety capability requirements of NFPA-805 and guidance of NEI 00-01.Reference Document EIR 51-9133191, NSCA, Section 8.0 EIR 51-9177678-000, Definitions Section EIR 32-9151404-000, Nine Mile Point Unit I -NFPA 805 Coordination Study NMPI, April 2013 Page B-57 I NMP1, April 2013 Page B-57 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref 3.3.1.1.4.1 Automatic Initiation Logics NEI 00-01 Guidance The automatic initiation logics for the credited post-fire safe shutdown systems are generally not required to support safe shutdown.Typically, each system can be controlled manually by operator actuation in the main control room or emergency control station. The emergency control station includes those plant locations where control devices, such as switches, are installed for the purpose of operating the equipment.

If operator actions to manually manipulate equipment at locations outside the MCR or the emergency control station are necessary, those actions must conform to the regulatory requirements on operator manual actions (See Appendix E). If not protected from the effects of fire, the fire-induced failure of automatic initiation logic circuits should be considered for their potential to adversely affect any post-fire safe shutdown system function.Applicability Applicable Alignment Statement Aligns Comments None Alignment Basis The Safe Shutdown Analysis takes credit for automatic transfer to an alternate power source if the transfer circuit and power source is not affected by the fire. As an example, the EC system can be initiated either manually or automatically.

The RPS instruments and logic that automatically initiate the EC system on high reactor pressure or low-low reactor level have been included in the analysis.

Manual initiation of the EC system can be accomplished from either the Control Room or RSP depending on the fire location.

AC power is not required to manually initiate DHR via the ECs.Adverse effects due to fire have been considered for the automatic initiation logic circuits.

Fire area compliance assessments demonstrate that safe shutdown capability is not adversely affected by a fire in any plant area that disables automatic functions (including initiation logic).Reference Document EIR 51-9133191, NSCA, Section 4.0 NMPI, April 2013 Page B-58 I I NMP1, April 2013 Page B-58 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref 3.3.1.1.5 Associated Circuits NEI 00-01 Guidance Cabling for the electrical distribution system is a concern for those breakers that feed 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 and classified as a required safe shutdown cable. 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.Additionally, the non-safe shutdown circuits off of each of the required safe shutdown components in the electrical distribution system can impact safe shutdown if not properly coordinated.

These cables are classified as required for hot shutdown based on the criteria contained in Appendix H.Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis The concern for cabling of the electrical distribution system for primary safe shutdown components involves those breakers that feed associated circuits and may not be fully coordinated with upstream breakers.

This involves circuits that are the direct power feed to primary safe shutdown components and/or the power feed to motor control centers that support components or other motor control centers. The concern for these circuits is not the load itself but the upstream power source.For the NMP1 electrical distribution system, it was assumed that the safe shutdown components and their associated power load were coordinated with their upstream power supplies when identifying cables for all existing and any new safe shutdown components.

For safe shutdown circuits, the cables included the direct power feed from the load center to the component and any cables associated with that component.

In addition, coordination was also assumed for any branch circuits related to the safe shutdown component's power source.Associated circuits are those circuits which are not completely independent of the safe shutdown systems or components.

Failure or spurious operation of these circuits could potentially defeat the safe shutdown capability of a safety system. A fire in a given fire area could potentially affect systems and components thought to be independent of that particular fire area.For the purpose of this analysis, an associated circuit must be associated with both a fire area and a safe shutdown system or component.

The associated circuits include circuits which share a common power supply with safe shutdown component and circuits whose spurious operation would adversely affect the safe shutdown capability.

Breaker coordination calculation 32-9151404-000, Nine Mile Point Unit 1 -NFPA 805 Coordination Study, demonstrates the existing coordination status (including any new electrical circuits) for required common power supplies that could impact safe shutdown.

This calculation identifies if there are any fire protection program breaker coordination issues concerning proper coordination between the supply breaker/fuse and the load breaker/fuses for power sources required for hot shutdown.

This document meets the nuclear safety capability requirements of NEI 00-01 and NFPA-805.NMPI, April 2013 Page B-59 I NMP1, April 2013 Page B-59 I Constellation Energy Nuclear Group Attachment B -P Reference Document EIR 51-9133191, NSCA, Section 8.5 EIR 32-9151404-000, Nine Mile Point Unit 1 -NFPA 805 Coordination Study JEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review I NMPI, April 2013 Page B-60 I I NMP1, April 2013 Page B-60 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref 3.3.1.1.6 Exclusion Analysis NEI 00-01 Guidance Exclusion analysis may be used to demonstrate a lack of potential for any impacts to post-fire safe shutdown from a component or group of components regardless of the cable routing. For these cases, rigorous cable searching and cable to component associations may not be required.Applicability Not Applicable Comments None Alignment Statement Not Required Alignment Basis Exclusion analysis was not used to demonstrate a lack of potential for any impacts to post-fire safe-shutdown.

Reference Document NMPI, April 2013 Page B-61 I!NMP1, April 2013 Page B-61 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref NEI 00-01 Guidance 3.3.2 Associated Circuit Of Concern Appendix R, through the guidance provided in NRC Generic Letter 81-12, requires that separation features be provided for associated Cables non-safety circuits that could prevent operation or cause mal-operation 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 7.1.5 and further clarified in a NRC memorandum dated March 22, 1982 from R. Mattson to D. Eisenhut, Reference 7.1.6. They are as follows:* Spurious actuations

  • Common power source* Common enclosure.

Each of these cables is classified as an associated circuit of concern cable.Cables Whose Failure May Cause Spurious Operations Safe shutdown system spurious operation concerns can result from fire damage to a cable whose failure could cause the spurious operation/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.

Common Power Source Cables The concern for the common power source associated circuits of concern 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.Common Enclosure Cables The concern with common enclosure associated circuits of concern 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.Applicability Comments Applicable None Alignment Statement Not Required Alignment Basis This section is a generic discussion concerning the separation of Associated Circuit Cables and non-safety circuits of components required for safe shutdown.

This information is discussed in more detail in subsequent sub-sections 3.3.3, 3.5.2.4 and 3.5.2.5.Reference Document I NMPI, April 2013 Page B-62 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref 3.3.3 Methodology for Cable Selection and Location NEI 00-01 Guidance 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 be circuits of concerns for a post-fire safe shutdown analysis.

Criteria for making the determination as to which circuits are to be classified as required for hot shutdown or as important to SSD is contained in Appendix H.Applicability Applicable Comments None Alignment Statement Not Required Alignment Basis This paragraph provides introductory information and contains no specific requirements.

Discussion is provided in subsequent sub-sections.

Reference Document NMPI, April 2013 Page B-63 I I NMPI, April 2013 Page B-63 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref NEI 00-01 Guidance 3.3.3.1 Identify Circuits Necessary For each piece of safe shutdown equipment defined in section 3.2, review the appropriate electrical diagrams including the following for the Operation of the Safe documentation to identify the circuits (power, control, instrumentation) required for operation or whose failure may impact the operation of Shutdown Equipment each piece of equipment:

  • Single-line electrical diagrams* Elementary wiring diagram* 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.Applicability Comments Applicable None Alignment Statement Aligns Alignment Basis Cable selection was performed to identify all conductors/wires that may be required for a component to perform its safe shutdown function, or whose failure could be adverse to the component's safe shutdown function.

The cables were selected by a point-to-point review of the applicable connection diagrams, single line electrical diagrams, elementary wiring diagrams or instrument loop wiring diagrams.During the initial review, any cable that had a potential to impact the safe shutdown component was identified and associated to that component.

Cables identified for each Safe Shutdown component, including any additional reference drawings, were populated in the database during the cable selection process. Cables that are computer inputs or that have adequate isolation are excluded.Figures 3-2 and 3-3 of NEI 00-01 were used to develop the Safe Shutdown Systems, the systems Paths and Safe Shutdown Equipment List. These lists included electrical distribution equipment identified for circuits whose failure may cause a coordination concern. The power related electrical equipment included upstream power sources up to either offsite power or the emergency power source.Coordination of power supplies was assumed when assigning cables to the safe shutdown components; however, this may not encompass any new components and circuits being added to the program. Breaker coordination calculation 32-9151404-000, Nine Mile Point Unit 1 -NFPA 805 Coordination Study, demonstrates the existing coordination status for the required common power sources. This calculation identifies if there are any fire protection program breaker coordination issues concerning proper coordination between the supply breaker/fuse and the load breaker/fuses for power sources required for hot shutdown.

This document meets the nuclear safety capability requirements of NEI 00-01, Section 3.5.2.4 and Figure 3.5.2-6, and NFPA 805, Section 2.4.2.2.2.

Reference Document EIR 51-9133191, NSCA, Sections 7.0 and 8.0 EIR 32-9151404-000, Nine Mile Point Unit 1 -NFPA 805 Coordination Study NMP1, April 2013 Page B-64 Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref NEI 00-01 Guidance 3.3.3.2 Identify Interlocked Circuits In reviewing each control circuit, investigate interlocks that may lead to additional circuit schemes, cables and equipment.

Assign to the and Cables that Could Affect Safe equipment any cables for interlocked circuits that can affect the equipment.

Shutdown 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 in an undesirable manner that impacts post-fire safe shutdown.Applicability Comments Applicable None Alignment Statement Aligns With Intent Alignment Basis For control logic circuits where multiple components receive signals from a common logic, the control logic was analyzed as a primary component and a pseudo-component identification was created for the control logic. A pseudo-component is an artificial association of equipment and cables that performs a common function into a single entity for analysis purposes.

The pseudo-component is not an actual plant hardware designation.

The pseudo-component was interlocked to the other associated primary components so that the effect of the control logic could be evaluated on an individual component level. This methodology was used for similar circuit scenarios such as common power supplies.

Whereas this approach does not assign the cable to each individual component, the effect on each component due to fire damage is analyzed.

This method serves to reduce the duplication of cable data when the same cables are assigned to multiple components.

Pseudo-components and other primary components whose associated cabling can affect another primary component based on interposing contacts were identified on the Cable Selection Worksheet of the affected component as an interlocked primary component.

This meets the intent of the guidance.Reference Document EIR 51-9133191, NSCA, Sections 5.4 and 6.0 EIR 51-9177678-&4=, Definitions Section NMPI, April 2013 Page B-65 I I NMP1, April 2013 Page B-65 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref 3.3.3.3 Assign Cables to Safe Shutdown Equipment NEI 00-01 Guidance Given the criteria/assumptions defined in Section 3.3.1, identify the cables required to operate or that may result in mal-operation of each piece of safe shutdown equipment.

Cables are classified as either required for hot shutdown or important to SSD based on the classification of the component to which they are associated and the function of that component in supporting post-fire safe shutdown in each particular fire area. Refer to Appendix H for additional guidance.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. Criteria for making the determination as to which cables are to be classified as required for hot shutdown or as important to SSD is contained in Appendix H.Applicability Applicable Alignment Statement Aligns Alignment Basis Comments None Cable selection for Safe Shutdown components was performed by identifying all cables required for a component to perform its safe shutdown function.

These cables were selected by a point-to-point review using the component's elementary diagram. The cables were selected in accordance with NEI 00-01, Section 3.3.1 and then entered into the database.

For cases where cables affected multiple components, either the cable was assigned to each component or a pseudo-component was used with the cables assigned to the pseudo-component instead of the primary component.

In addition to the cables, any component interlocks were identified to investigate their impact on the operation of the safe shutdown component.

The relationship between these interlocks and the primary component were documented and their dependency was considered during the Fire Area Compliance Assessment.

Coordination of power supplies is addressed in Section 3.5.2.4 of this document.Reference Document EIR 51-9133191, NSCA, Section 8.0 EIR 51-9177678-=0, Definitions Section NMPI, April 2013 Page B-66 I!NMP1, April 2013 Page B-66 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref 3.5 Circuit Analysis and Evaluation NEI 00-01 Guidance This section on circuit analysis provides information on the potential impact of fire on circuits used to monitor, control and power required for hot shutdown and important to 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.

Additionally, when assessing fire-induced damage to circuits that could potentially result in MSOs, the circuit failure criteria in Appendix B should be used.Appendix R Section II.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.Applicability Applicable Comments None Alignment Statement Not Required Alignment Basis This paragraph provides introductory information and contains no specific requirements.

Discussion is provided in subsequent sub-sections.

Reference Document NMP1, April 2013 Page B-67 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref NEI 00-01 Guidance 3.5.1 Criteria/Assumptions Apply the following criteria/assumptions when performing fire-induced circuit failure evaluations.

Refer to the assessment of the NEI/EPRI and CAROLFIRE Cable Test Results in Appendix B to this document for the basis for these criteria and for further elaboration on the application of the criteria.Applicability Comments Applicable None Alignment Statement Not Required Alignment Basis This paragraph provides introductory information and contains no specific requirements.

Discussion is provided in subsequent sub-sections.

Reference Document NM P1, April 2013 Page B-68 I I NMP1, April 2013 Page B-68 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.1 Circuit Failure Criteria Circuit Failure Criteria:

The criteria provided below addresses the effects of multiple fire-induced circuit failures impacting circuits for components classified as either "required for hot shutdown" or "important to safe shutdown".

Consider the following circuit failure types on each conductor of each unprotected cable. Criteria differences, however, do apply depending on whether the component is classified as required for hot shutdown or important to safe shutdown." 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.

0 A hot short in the control circuitry for an MOV can bypass the MOV protective devices, i.e. torque and limit switches.

This is the condition described in NRC Information Notice 92-18. In this condition, the potential exists to damage the MOV motor and/or valve. Damage to the MOV could result in an inability to operate the MOV either remotely, using separate controls with separate control power, or manually using the MOV hand wheel. This condition could be a concern in two instances:

(1) For fires requiring Control Room evacuation and remote operation from the Remote Shutdown Panel, the Auxiliary Control Panel or Auxiliary Shutdown Panel; (2) For fires where the selected means of addressing the effects of fire induced damage is the use of an operator manual action. In each case, analysis must be performed to demonstrate that the MOV can be subsequently operated electrically or manually, as required by the safe shutdown analysis.* 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: Open circuits as a result of conductor melting have not occurred in any of the recent cable fire testing and they are not considered to be a viable form of cable failure.]" 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. A short-to-ground may also result in a change of state for normally energized equipment.

Circuits for "required for hot shutdown" components:

Because Appendix R Section III.G.1 requires that the hot shutdown capability remain"free of fire damage", there is no limit on the number of concurrent/simultaneous fire-induced circuit failures that must be considered for circuits for components "required for hot shutdown:

located within the same fire area. For components classified as "required for hot shutdown", there is no limit on the duration of the hot short. It must be assumed to exist until an action is taken to mitigate its effects.Circuits required for the operation of or that can cause the mal-operation of "required for hot shutdown" components that are impacted by a fire are considered to render the component unavailable for performing its hot shutdown function unless these circuits are properly protected as described in the next sentence.

The required circuits for any "required for hot shutdown" component, if located within the same fire area where they are credited for achieving hot shutdown, must be protected in accordance with one of the requirements of Appendix R Section III.G.2 or plant specific license conditions.

NMPI, April 2013 Page B-69 I I NMP1, April 2013 Page B-69 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Circuits for "important to safe shutdown" components:

Circuits for components classified as "important to safe shutdown" are not specifically governed by the requirements of Appendix R Section III.G.1, III.G.2 or III.G.3. To address fire-induced impacts on these circuits, consider the three types of circuit failures identified above to occur individually on each conductor with the potential to impact any"important to safe shutdown" component with the potential to impact components "required for hot shutdown".

In addition, consider the following additional circuit failure criteria for circuits for "important to safe shutdown" components located within the same fire area with the potential to impact components "required for hot shutdown": " As explained in Figure 3.5.2-3, multiple shorts-to-ground are to be evaluated for their impact on ungrounded circuits.* As explained in Figure 3.5.2-5, for ungrounded DC circuits, a single hot short from the same source is assumed to occur unless it can be demonstrated that the occurrence of a same source short is not possible in the affected fire area. If this approach is used, a means to configuration control this condition must be developed and maintained." For the double DC break solenoid circuit design discussed in the NRC Memo from Gary Holahan, Deputy Director Division of Systems Technology, dated December 4, 1990 and filed under ML062300013, the effect of two hot shorts of the proper polarity in the same multi-conductor cable should be analyzed for non-high low pressure interface components.

[Reference Figure B.3.3 (f) of NFPA 805-2001.]" Multiple spurious operations resulting from a fire-induced circuit failure affecting a single conductor must be included in the post-fire safe shutdown analysis.* Multiple fire-induced circuit failures affecting multiple conductors within the same multi-conductor cable with the potential to cause a spurious operation of an "important to safe shutdown" component must be assumed to exist concurrently." Multiple fire-induced circuit failures affecting separate conductors in separate cables with the potential to cause a spurious operation of an "important to safe shutdown" component must be assumed to exist concurrently when the effect of the fire-induced circuit failure is sealed-in or latched.Conversely, multiple fire-induced circuit failures affecting separate conductors in separate cables with the potential to cause a spurious operation of an "important to safe shutdown" component need not be assumed to exist concurrently when the effect of the fire-induced circuit failure is not sealed-in or latched. This criterion applies to consideration of concurrent hot shorts in secondary circuits and to their effect on a components primary control circuit. It is not to be applied to concurrent single hot shorts in primary control circuit for separate components in an MSO combination.

  • For components classified as "important to safe shutdown", the duration of a hot short may be limited to 20 minutes. (If the effect of the spurious actuation involves a "sealing in" or "latching" mechanism, that is addressed separately from the duration of the spurious actuation, as discussed above.)* For any impacted circuits for "important to safe shutdown" components that are located within the same fire area, protection in accordance with the requirements of Appendix R Section Ill.G.2 or plant specific license conditions may be used. In addition, consideration may be given to the use of fire modeling or operator manual actions, as an alternative to the requirements of Appendix R Section III.G.2. (Other resolution options may also be acceptable, if accepted by the Authority Having Jurisdiction.)

NMPI, April 2013 Page B-70 IlNMP1, April 2013 Page B-70 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Applicability Comments Applicable None Alignment Statement Aligns Alignment Basis During the cable selection process, a circuit fault analysis for each component cable was initially performed to determine the effects of a fire-induced hot short, open circuit and short to ground, as applicable.

Per the NEI 00-01 guidance, all combinations of circuit failures (hot shorts, open circuit, and short-to ground) on each conductor for each unprotected safe shutdown cable were considered.

Further analysis was performed, as required, for secondary or interlocked circuits.The circuit failures were evaluated to determine the potential impact of a fire on the safe shutdown equipment (including the path) that is associated with that cable/conductor.

In some cases, the cables and components had adequate separation from their redundant circuits and components as required by the regulations and were not required to be analyzed.It was assumed that fire damage results in an unusable cable that cannot be considered functional with regard to ensuring proper circuit operation.

The insulation and external jacket material of electrical cables are susceptible to fire damage. Damage may assume several forms including deformation, loss of structure, cracking, and ignition.

The relationship between exposure of electrical cable insulation to fire conditions, the failure mode, and time to failure may vary with the configuration and cable type. To accommodate these uncertainties in a consistent and conservative manner, the circuit analysis assumes that the functional integrity of electrical cables is lost when cables are exposed to a fire, except where protected by a fire rated barrier.The types of circuit failures considered for this analysis are those identified in NEI 00-01, Appendix B, Table B.1.0, "Types of Fire-Induced Circuit Failures Required to Be Considered." Consistent with NEI 00-01, hot shorts were considered to be either internal cable wire-to-wire shorts or external cable-to-cable shorts. No credit was taken for physical cable attributes (armored, thermo-set, etc.) preventing cable-to-cable hot shorts.Reference Document EIR 51-9133191, NSCA, Section 8.2 HNP RAI 3-16, NRC Request for Addition Information (ML092170715)

NMPI, April 2013 Page B-71 I!NMP1, April 2013 Page B-71 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.2 Spurious Operation Criteria Spurious Operation Criteria:

The following criteria address the effect of multiple spurious operations of components classified as either"required for hot shutdown" or "important to safe shutdown" on post-fire safe shutdown.

These criteria are to be applied to the population of components whose spurious operation has been determined to be possible based on an application of the circuit failure criteria described above when assessing impacts to post-fire safe shutdown capability in any fire area.* The set of concurrent combinations of spurious operations provided through the MSO Process outlined in Section 4 and the list of MSO contained in Appendix G must be included in the analysis of MSOs.* MSOs do not need to be combined, except as explained in Section 4.4.3.4 of this document.* Section 4.4.3.4 states that the expert panel should review the plant specific list of MSOs to determine whether any of the individual MSOs should be combined due to the combined MSO resulting in a condition significantly worse than either MSO individually.

  • In this review, consideration of key aspects of the MSOs should be factored in, such as the overall number of spurious operations in the combined MSOs, the circuit attributes in Appendix B, and other physical attributes of the scenarios.

o Specifically, if the combined MSOs involve more than a total of four components or if the MSO scenario requires consideration of sequentially selected cable faults of a prescribed type, at a prescribed time, in a prescribed sequence in order for the postulated MSO combination to occur, then this is considered to be beyond the required design basis for MSOs.Applicability Comments Applicable None Alignment Statement Aligns Alignment Basis An MSO Expert panel reviewed the generic list of scenarios listed in NEI-00-01 Appendix G, screening out any scenario that was a combination of five or more spurious operations.

There were no screening criteria based on number, timing, or type of circuit failures.

The result of the review was a list of MSO scenarios, both generic and site specific, that were included in the Fire Area Analysis for further evaluation.

Reference Document EIR 51-9133191, NSCA, Section 8.1 ONS RAI 3-38, NRC Request for Additional Information dated July 30, 2010 (ML102110394)

Technical Report on Identification

& Classification of the NMP-1 MSO Scenarios Using an Expert Panel -Review of New Scenarios, Rev.1 NMPI, April 2013 Page B-72 I I NMPI, April 2013 Page B-72 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref 3.5.1.3 Circuit Contact Position NEI 00-01 Guidance Assume that circuit contacts are initially positioned (i.e., open or closed) consistent with the normal mode/position of the "required for hot shutdown" or "important to safe shutdown" equipment as shown on the schematic drawings.

The analyst must consider the position of the"required for hot shutdown" and "important to 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 "required for hot shutdown" and "important to safe shutdown equipment".

Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis The analysis assumes that the circuit contacts are positioned (i.e., open or closed) consistent in the normal mode/position of the safe shutdown equipment as shown on the schematic drawings or defined by procedure.

The fire damage impact on the position of the safe shutdown equipment was considered for each shutdown scenario.Reference Document EIR 51-9133191, NSCA, Section 8.6.1 NMPI, April 2013 Page B-73 I NMP1, April 2013 Page B-73 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref 3.5.2 Types Of Circuit Failures NEI 00-01 Guidance 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 required for hot shutdown and important to safe 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 necessary to achieve and maintain 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 hot shutdown be free of fire damage and that these circuits be designed for the fire-induced effects of a hot short, short-to-ground, or an open circuit. With respect to the electrical distribution system, the issue of breaker coordination must also be addressed.

Criteria for making the determination as to which breakers are to be classified as required for hot shutdown is contained in Appendix H.This section will discuss specific examples of each of the following types of circuit failures:* Open circuit* Short-to-ground

  • Hot short Also, refer to Appendix B for the circuit failure criteria to be applied in assessing the impact of the Plant Specific List of MSOs on post-fire safe shutdown.Applicability Applicable Comments None Alignment Statement Not Required Alignment Basis This paragraph provides introductory information and contains no specific requirements.

Discussion is provided in subsequent sub-sections.

Reference Document NMPI, April 2013 Page B-74 I I NMPI, April 2013 Page B-74 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.1 Circuit Failures Due to an This section provides guidance for addressing the effects of an open circuit for required for hot shutdown and important to safe shutdown Open Circuit 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.* Loss of electrical continuity may occur within a conductor resulting in de-energizing the circuit and causing a loss of power to, or control of, the required for hot shutdown and important to 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, possibly resulting in the occurrence of an additional fire in the location of the CT itself.Applicability Comments Applicable None Alignment Statement Aligns Alignment Basis Open circuits are analyzed as referenced in the NEI 00-01 guidance, Section 3.5.2.1 and Figure 3.5.2-1. An open circuit is a condition experienced when an individual conductor within a cable loses electrical continuity due to a fire induced break. This could cause the loss of power from de-energizing the circuit or the ability to control affected components, or on energized equipment could cause a change of position of the component.

In addition, as stated in the guidance, an open circuit on a high voltage ammeter CT circuit may result in secondary damage to that circuit.The Nuclear Safety Capability Assessment (NSCA) assumed that fire damage results in an unusable cable that cannot be considered functional with regard to ensuring proper circuit operation.

The insulation and external jacket material of electrical cables are susceptible to fire damage. Damage may assume several forms including deformation, loss of structure, cracking, and ignition.

The relationship between exposure of electrical cable insulation to fire conditions, the failure mode, and time to failure may vary with the configuration and cable type. To accommodate these uncertainties in a consistent and conservative manner, the circuit analysis assumes that the functional integrity of electrical cables was lost when cables are exposed to a fire, except where protected by a fire rated barrier.Other associated circuit concerns are related to open secondary circuits and 4 KV bus CT that may result in high currents producing an additional fire at the transformer location.

This associated circuit concern is related to CT where an open secondary circuit may develop high voltages within a transformer potentially resulting in a secondary fire at the transformer location.

This issue is evaluated in Fire Protection Engineering Evaluation FPEE-1-04-002, Rev. 0, Fire Effects on CTs and Instrument Sensing Lines and The Plant Safe Shutdown Capability.

The evaluation concludes that, for all CTs in use at NMP1, an open transformer secondary will not develop voltages that are high enough to threaten the Safe Shutdown capability.

Reference Document EIR 51-9133191, NSCA, Sections 8.2 and 8.5 Fire Protection Engineering Evaluation FPEE-1 002, Rev. 0, Fire Effects on CTs ONS RAI 3-48, NRC Request for Additional Information dated July 30, 2010 (ML102110394)

HNP RAI 3-17, NRC Request for Additional Information dated August 6, 2009 (ML092170715)

I NMP1, April 2013 Page B-75 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref 3.5.2.2 Circuit Failures Due to a Short-to-Ground NEI 00-01 Guidance This section provides guidance for addressing the effects of a short-to-ground on circuits for required for hot shutdown and important to 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.There is no limit to the number of shorts-to-ground that could be caused by the fire.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.

Comments None Applicability Applicable Alignment Statement Aligns Alignment Basis The methodology assumes multiple fire induced failures including short-to-ground.

The short-to-ground issue incorporates circuit failures for both ungrounded and grounded circuits.Postulated cable and component failures were identified utilizing the techniques referenced in NEI 00-01 Figure 3.5.2-2 and Figure 3.5.2-3. The safe shutdown analysis may exclude certain cables if their postulated fire induced faults have no adverse effect on the component.

A short-to-ground fault for grounded circuits could cause the tripping of a circuit thereby causing a loss of power to the control circuit. For certain cases of energized components, a loss of control power may result in a loss of power to relays and other devices interlocked with the device.Unless otherwise justified by circuit analysis, short- to-ground for ungrounded circuits are treated the same as short- to-ground for grounded circuits, and are postulated to result in a loss of motive power or control power. This is consistent with NFPA 805 Appendix B, Section B.3.2.g which states: "For ease of analysis when analyzing an ungrounded DC circuit for the effects of a short-to-ground, it should be assumed that an existing ground fault from the same power source is present." Reference Document EIR 51-9133191, NSCA, Sections 8.1 and 8.2 NFPA 805, Appendix B, Section B.3.2.g NMP1, April 2013 Page B-76 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.3 Circuit Failures Due to a Hot This section provides guidance for analyzing the effects of a hot short on circuits for required for required for hot shutdown and important Short to 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 operation of equipment.

The spuriously operated device (e.g., relay) may be interlocked with another circuit that causes the spurious operation of other equipment.

This type of hot short is called an intra-cable hot short (also known as 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 and a de-energized conductor may also cause a spurious operation of equipment.

This is called an inter-cable hot short (also known as cable-to-cable hot short/external hot short).* A hot short in the control circuitry for an MOV can bypass the MOV protective devices, i.e. torque and limit switches.

This is the condition described in NRC Information Notice 92-18. In this condition, MOV motor damage can occur. Damage to the MOV motor could result in an inability to operate the MOV either remotely, using separate controls with separate control power, or manually using the MOV hand wheel. This condition could be a concern in two instances:

(1) For fires requiring Control Room evacuation and remote operation from the Remote Shutdown Panel; (2) For fires where the selected means of addressing the effects of fire induced damage is the use of an operator manual action. In this latter case, analysis must be performed to demonstrate that the MOV thrust at motor failure does not exceed the capacity of the MOV hand wheel. For either case, analysis must demonstrate the MOV thrust at motor failure does not damage the MOV pressure boundary.Applicability Comments Applicable None Alignment Statement Aligns Alignment Basis A hot short is a condition experienced when an energized individual conductor of the same or different cable comes into contact with another conductor of the same or different cable resulting in electrical continuity between the conductors.

The potential effect is that the energized conductor becomes an undesired source of power for the circuit being analyzed.

Hot shorts were considered to be either internal conductors of the same cable, identified as internal shorts, or shorts between conductors of different cables, identified as external shorts.The potential of circuit failures due to hot shorts can cause components to operate or cause them to fail to operate in an undesired manner.The Nuclear Safety Capability Assessment (NSCA) assumed that fire damage results in an unusable cable that cannot be considered functional with regard to ensuring proper circuit operation.

The insulation and external jacket material of electrical cables are susceptible to fire damage. Damage may assume several forms including deformation, loss of structure, cracking, and ignition.

The relationship between exposure of electrical cable insulation to fire conditions, the failure mode, and time to failure may vary with the configuration and cable type. To accommodate these uncertainties in a consistent and conservative manner, the circuit analysis assumes that the functional integrity of electrical cables is lost when cables are exposed to a fire, except where protected by a fire rated barrier.NMPI, April 2013 Page B-77 I I NMP1, April 2013 Page B-77 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Consistent with NEI 00-01, hot shorts were considered to be either internal cable wire-to-wire shorts or cable-to-cable (external) shorts. No credit was taken for physical cable attributes (armored, thermo-set, etc.) preventing cable-to-cable hot shorts.For cable failures due to hot shorts on grounded or ungrounded circuits, the methodology initially assumes the hot short would have sufficient potential to cause a spurious operation of the component.

Two types of cable hot short conditions are considered to be of sufficiently low likelihood that they are not assumed credible, except for analysis involving high/low pressure interface components.

These hot shorts are 3-phase AC power circuit cable-to-cable proper phase sequence faults and 2-wire ungrounded DC circuit cable-to-cable proper polarity faults.Instrument circuits that operate at low signal levels (4-20 mA, 0-1 V, 1-5 V, etc.) and are enclosed in a grounded metal shield are not considered to be susceptible to hot shorts from other adjacent instrument circuits external to the shield. External circuits are assumed to short to ground via the shield and do not have the potential of creating a signal of proper polarity and amplitude to simulate a valid instrument signal.Reference Document EIR 51-9133191, NSCA, Section 8.2 ONS RAI 3-41, NRC Request for Additional Information dated July 30, 2010 (ML102110394)

NMPI, April 2013 Page B-78 I NMP1, April 2013 Page B-78 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.3 Nuclear Safety Equipment and Cable Location Physical location of equipment and cables shall be identified.

NEI 00-01 Ref 3.3.3.4 Identify Routing of Cables NEI 00-01 Guidance 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.Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis This task involved identifying the routing and location for all raceways and endpoints for cables associated with safe shutdown equipment.

The cable routes and their endpoint location were populated into the database.

The database is a relational database that contains all the required information for safe shutdown cable routing and endpoint information.

The original cable routing and cable endpoint data was provided from the NMP1 cable raceway database (TRAK2000).

Comments None Reference Document EIR 51-9133191, NSCA, Sections 2.1 and 8.4 TRAK2000, Revision 6.01 NMPI, April 2013 Page B-79 I I NMPI, April 2013 Page B-79 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.3 Nuclear Safety Equipment and Cable Location NEI 00-01 Ref 3.3.3.5 Identify Raceway and Cables by Fire Area NEI 00-01 Guidance 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. For raceway and cable endpoints in multiple fire areas, each fire area where the raceway or cable endpoint exists must be included.

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.Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis Fire area locations were identified for each cable raceway and cable endpoint by obtaining the location coordinates from applicable cable tray, conduit or equipment layout arrangement drawings or by field walkdown, if necessary.

The fire area locations were identified by comparing the cable tray/conduit arrangement drawings, equipment arrangement drawings, or field walkdown data with Fire Area Floor Plans drawings.

This correlation between cable raceway locations and Fire Areas and Rooms was populated into the database to produce computer generated reports. The reports contained the cable related raceway information required to prepare the Fire Area Analysis.Reference Document EIR 51-9133191, NSCA, Sections 5.4 and 8.4 Fire Area Floor Plans Drawings B40141C through B40148C NMPI, April 2013 Page B-80 I!NMP1, April 2013 Page B-80 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.3 Nuclear Safety Equipment and Cable Location NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.4 Circuit Failures Due to The evaluation of circuits of a common power source consists of verifying proper coordination between the supply breaker/fuse and the Inadequate Circuit Coordination load breakers/fuses for power sources that are required for hot 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.

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.The methodology for identifying potential circuits of a common power source and evaluating circuit coordination cases 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 example, for breakers that have internal breaker tripping devices and do not require control power to trip the breaker, assure that the time-current characteristic curve for any affected load breaker is to the left of the time-current characteristic curve for the bus feeder breaker and that the available short circuit current for each affected breaker is to the right of the time-current characteristic curve for the bus feeder breaker or that the bus feeder breaker has a longer time delay in the breaker instantaneous range than the load breaker. For breakers requiring control power for the breaker to trip, the availability of the required control power must be demonstrated in addition to the proper alignment of the time-current characteristic curves described above. The requirement for the availability of control power would apply to load breakers fed from each safe shutdown bus where a fire-induced circuit failure brings into questions the availability of coordination for a required for hot shutdown component.
  • 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 circuit of concern are routed and the power source is required for hot shutdown.

Prepare a list of the following information for each fire area:* Cables of concern.* 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 circuit of concern cables routed in an area of the same path as required by the power source. Evaluate adequate separation and other mitigation measures based upon the criteria in Appendix R, NRC staff guidance, and plant licensing bases.NMPI, April 2013 Page B-81 I NMP1, April 2013 Page B-81 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Applicability Comments Applicable None Alignment Statement Aligns Alignment Basis The NMP1 Fire Area Assessments are performed to support transition to a performance based fire protection licensing basis. While performing the primary component circuit analysis for safe shutdown components, it was assumed that electrical coordination exists for all power supplies for each level of electrical power.Breaker coordination calculation 32-9151404-000, Nine Mile Point Unit 1 -NFPA 805 Coordination Study, demonstrates the existing coordination status for the required common power sources. This calculation identifies any fire protection program breaker coordination issues concerning proper coordination between the supply breaker/fuse and the load breaker/fuses for power sources required for hot shutdown.

This document meets the nuclear safety capability requirements of NEI 00-01, Section 3.5.2.4, Figure 3.5.2-6 and NFPA 805, Section 2.4.2.2.2.

Any identified issues have been addressed in the NSCA.Other potential coordination concerns involve associated non-safe shutdown circuits that are not independent of safe shutdown circuits that could also potentially defeat the functions of safe shutdown circuits if not properly protected.

These circuits must be associated with both a fire area and a safe shutdown system or component to warrant consideration.

These associated circuits are divided into three categories.

  • Circuits that share a common power supply with safe shutdown circuits.* Circuits that share a common enclosure with safe shutdown circuits.* Circuits for components the spurious operation of which would adversely affect the shutdown process.The associated circuits are defined in the NMP1 Coordination study which reviews the 4.16 kV, 600 VAC, 480 VAC, 208/120 VAC and 125 VDC power supplies credited for post-fire shutdown.Proper circuit coordination for power supplies was reviewed, analyzed and addressed in EIR 51-9133191, NSCA.Reference Document EIR 51-9133191, NSCA, Section 8.5 HNP RAI 3-18 and RAI 3-19, NRC Request for Additional Information dated August 6, 2009 (ML092170715)

EIR 32-9151404-000, Nine Mile Point Unit 1 -NFPA 805 Coordination Study NMPI, April 2013 Page B-82 I I NMP1, April 2013 Page B-82 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.3 Nuclear Safety Equipment and Cable Location NEI 00-01 Ref 3.5.2.5 Circuit Failures Due to Common Enclosure Concerns NEI 00-01 Guidance The common enclosure concern deals with the possibility of causing 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.Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis Circuit failures due to common enclosure concerns are addressed by breaker coordination calculation 32-9151404-000, Nine Mile Point Unit 1 -NFPA 805 Coordination Study. This calculation demonstrates the existing coordination status for electrical circuits that could impact safe shutdown concerning proper coordination between the supply breaker/fuse and the load breaker/fuses for power sources required for hot shutdown.

This document meets the nuclear safety capability requirements of NEI 00-01, Section 3.5.2.4, Figure 3.5.2-6 and NFPA 805, Section 2.4.2.2.2.

Reference Document EIR 51-9133191, NSCA, Section 8.5 HNP RAI 3-18 and RAI 3-19, NRC Request for Additional Information dated August 6, 2009 (ML092170715)

EIR 32-9151404-000, Nine Mile Point Unit 1 -NFPA 805 Coordination Study NMPI, April 2013 Page B-83 IlNMP1, April 2013 Page B-83 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment 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).

NEI 00-01 Ref 3.4 Fire Area Assessment and Compliance Strategies NEI 00-01 Guidance 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.

Knowing which components and systems are performing which safe shutdown functions, the required and important to SSD components can be classified.

Once these component classifications have been made the tools available for mitigating the effects of fire induced damage can be selected.Refer to Appendix H for additional guidance on classifying components as either required for hot shutdown or important to safe shutdown.For MSOs the Resolution Methodology outlined in Section 4, Section 5, Appendix B and Appendix G should be applied. Components in each MSO are classified as either required for hot shutdown or important to safe shutdown components using the criteria from Appendix H. Similarly, this classification determines the available tools for mitigating the effects of fire-induced damage to the circuits for these components.

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.Applicability Applicable Comments None Alignment Statement Not Required Alignment Basis This paragraph provides introductory information and contains no specific requirements.

Discussion is provided in subsequent sub-sections.

Reference Document NMPI, April 2013 Page B-84 I I NMP1, April 2013 Page B-84 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref 3.4.1 Criteria/Assumptions NEI 00-01 Guidance The following criteria and assumptions apply when performing "deterministic" 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.Applicability Applicable Comments None Alignment Statement Not Required Alignment Basis This paragraph provides introductory information and contains no specific requirements.

Discussion is provided in subsequent sub-sections.

Reference Document NMPI, April 2013 Page B-85 I I NMP1, April 2013 Page B-85 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref 3.4.1.1 Assume a Single Fire NEI 00-01 Guidance Assume only one fire in any single fire area at a time.Applicability Comments Applicable None Alignment Statement Aligns Alignment Basis Only one fire is assumed to occur in any single fire area at a time.Reference Document EIR 51-9133191, NSCA, Section 9.0 NMPI, April 2013 Page B-86 I I NMP1, April 2013 Page B-86 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref 3.4.1.2 Fire Affects All Unprotected Cables and Equipment NEI 00-01 Guidance 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 postulated in the regulation.

Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis For a conservative approach which bounds the exposure fire required by the regulations, the analysis assumes a fully involved fire and that all equipment and unprotected cabling within a given fire area are damaged by the fire.Reference Document EIR 51-9133191, NSCA, Section 9.0 NMPI, April 2013 Page B-87 I!NMP1, April 2013 Page B-87 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref 3.4.1.3 Address all Cable and Equipment Impacts Affecting the Required Safe Shutdown Path NEI 00-01 Guidance 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.

Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis Fire Area Assessments were performed on a Fire Area basis in order to ensure compliance in accordance with the safe shutdown requirements of NFPA 805. The Safe Shutdown System and Component drawings were analyzed for each Fire Area to ensure that a success path is available based upon the postulated equipment and/or cable losses in the area.The potentially affected equipment and cables in each Fire Area were reviewed and impacts on safe shutdown success paths analyzed.The route and location of all safe shutdown cables were loaded into the safe shutdown database.

This data was used to generate Fire Area Component Impact Reports, which identified affected systems and components on a Fire Area basis. The Fire Area Component Impact Reports were used as a means to determine the least impacted safe shutdown path for each fire area.The Fire Area Analysis methodology assumed multiple fire-induced failures and multiple spurious actuations, based on the cables and components present in the fire area of concern.All postulated cable and component failures were identified and a resolution provided at the component level.The least impacted safe shutdown success path was analyzed so that mitigating strategies could be developed and documented in the Fire Area Assessment.

A success path determination for all safe shutdown functions was performed.

Generally the path with the least amount of failures was recovered to demonstrate a success path for safe shutdown.Support systems were reviewed in order to assess the impact on the systems being supported.

The credited safe shutdown success path was documented in the fire area compliance assessment.

Reference Document EIR 51-9133191, NSCA, Section 9.0 NMPI, April 2013 Page B-88 I I NMP1, April 2013 Page B-88 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref 3.4.1.4 Classify Each Cable/Component NEI 00-01 Guidance Use the criteria from Appendix H to classify each impacted cable/component as either a required or important to SSD cable/component.

Applicability Applicable Comments None Alignment Statement Not Required Alignment Basis Using the criteria from Appendix H to classify each impacted cable/component as either a required or important to SSD cable/component was not required to support the transition to NFPA 805. Therefore, cables and components were not classified as "required for safe shutdown" or "important for safe shutdown." The safe shutdown flow paths identify the primary components that are required to meet the safe shutdown performance goals. The safe shutdown cables/components were compiled based on each system's performance and safe shutdown function.

These components establish the primary safe shutdown flowpath for system operation.

Also included in the safe shutdown flow paths are those cables/components whose spurious operation could impact safe shutdown system operability.

Systems, components, and cables identified as necessary for the operation of the safe shutdown system under review are included in the safe shutdown equipment and cables lists and are designated with the same shutdown path as the primary safe shutdown system.The components may involve branch flow paths that must be isolated and remain isolated to assure that flow will not be diverted from the primary flow path. The list of primary components may also include selected mechanical components required to support safe shutdown.Reference Document EIR 51-9133191, NSCA, Section 9.0 NMPI, April 2013 Page B-89 I!NMP1, April 2013 Page B-89 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref 3.4.1.5 Manual Actions NEI 00-01 Guidance Use operator manual actions where appropriate, for cable/component impacts classified as important to SSD cable/components, to achieve and maintain post-fire safe shutdown conditions in accordance with NRC requirements (refer to Appendix E). For additional criteria to be used when determining whether an operator manual action may be used for a flow diversion off of the primary flow path, refer to Appendix H.Applicability Applicable Alignment Statement Aligns Comments None Alignment Basis Manual actions performed as prescribed in procedures or otherwise are documented in each Fire Area Assessment (FAA). Included in each FAA was the identification of any required operator actions outside the Main Control Room. The operator actions are those directed by operating procedures, repair procedures, or otherwise identified as necessary during the course of the individual FAA. Actions performed at locations other than primary control stations are identified as recovery actions requiring further review as part of the fire risk evaluations.

These actions are identified in the Report of Manual Actions and Report of Procedure Directed Manual Actions included in each FAA.Reference Document EIR 51-9133191, NSCA, Section 9.0 NMPI, April 2013 Page B-90 I I NMP1, April 2013 Page B-90 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref 3.4.1.6 Repairs NEI 00-01 Guidance 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 to equipment required in support of post-fire shutdown.Applicability Applicable Alignment Statement Comments None Aligns with Intent Alignment Basis Repairs whito are relied upon t achieve and maintain cold shutdown will be performed as Fequired.

Repairs are directed by plant procedures.

However, NFPA 805 requires only that the plant be maintained in a safe and stable condition.

Nor does NFPA 805 require that the plant achieve cold shutdown in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Refer to Section 4.2.1.2 for a description of safe and stable as applied to NMP1. NMPI demonstrates the ability to maintain for each fire area the fuel in a safe and stable condition with one of four designated shutdown paths.Reference Document EIR 51-9133191, NSCA, Sections 2.1, 8.5, and 9.0 Closure of National Fire Protection Association 805 Transition Program Frequently Asked Question Number 08-0054 (MLI 10140183)NMPI, April 2013 Page B-91 I!NMP1, April 2013 Page B-91 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.7 Appendix R Compliance For the components on the required safe shutdown path classified as required hot shutdown components as defined in Appendix H, Criteria 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 (III.G.1 .a). When cables or equipment 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 components impacted circuit(s): " Separation of cables and equipment and associated non-safety circuits of redundant trains within the same fire area by a fire barrier having a 3-hour rating (lll.G.2.a)" Separation of cables and equipment and associated non-safety 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 (lll.G.2.c).

For fire areas inside non-inerted containments, the following additional options are also available:

  • Separation of cables and equipment and associated non-safety circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards (llI.G.2.d);
  • Installation of fire detectors and an automatic fire suppression system in the fire area (llI.G.2.e);
  • 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, LARs and licensing change processes to satisfy the requirements mentioned above and to demonstrate equivalency depending upon the plant's license requirements.

Applicability Comments Applicable None Alignment Statement Aligns With Intent Alignment Basis Each NMP1 fire area containing safe shutdown equipment or cables was reviewed in a deterministic fashion for the ability to achieve post-fire safe shutdown.

The affected shutdown related cables and components in each area were identified and the resultant information used to determine the preferred shutdown path to achieve safe shutdown.The credited safe shutdown success paths were analyzed and mitigating strategies (procedu"al

...i.n., .epair actios.. Or ..

were developed and documented in the Safe Shutdown Analysis, fire area compliance assessments.

The results of the assessments confirm that in the event of a postulated exposure fire, the safe shutdown capability of NMP1 will be maintained such that the fuel remains in a safe and stable condition..

NMPI, April 2013 Page B-92 I I NMP1, April 2013 Page B-92 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review The non-safe shutdown circuits which are not completely independent of safe shutdown circuits are associated circuits.

These associated circuits are those circuits that could adversely affect the safe shutdown capability or components.

These circuits would be associated with a safe shutdown system or component and analyzed the same as other safe shutdown circuits affected within that fire area.Reference Document EIR 51-9133191, NSCA, Sections 5.1, 5.2, and 5.3 NMPI, April 2013 Page B-93 I I NMPI, April 2013 Page B-93 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.8 Alternate/Backup Equipment Consider selecting other equipment that can perform the same safe shutdown function as the impacted equipment.

In addressing this Selection situation, each equipment impact, including spurious operation, is to be addressed in accordance with regulatory requirements and the NPP's current licensing basis. With respect to MSOs, the criteria in Section 4, Appendix B, Appendix G and Appendix H should be used.Applicability Comments Applicable Consideration of Multiple Spurious Operations is addressed in the NSCA (EIR 51-9133191, Section 8.6).Alignment Statement Aligns Alignment Basis Component selection was performed for all fire areas in order to populate the database with equipment information required to be analyzed against the requirements of NFPA 805.10CFR50, Appendix R. The components selected are documented in the Safe Shutdown Equipment List (SSEL). The objective of the SSEL is to provide a list of analyzed components that are utilized in the NFPA 805 NSCA to demonstrate the fuel can be maintained in safe and stable condition, post firo safo shutdwn analysis to .ncuro that (1) one succesrs path (turutu Fr.s, system s, and rnmpo n ents) noceessarFy to aih ioo w-sA-afe s hut-do- ie' 46 frog o1Gf firoe dam ag a without red itinRg plant Or systemA repair F apab 0iities and (2 ) one succes6 path (sVUtructuo, systems, and components) necessary to adhieve Gcold- s-hutd-GHow

%it-hin 72 haoura is free of fira damage, Or Awaiqlabl

Athin 72 hour-s after Grediting plant Or system ropair -apabilities-.The current SSEL was reviewed against the criteria outlined in NEI 00-01 and considered where additional equipment may need to be included to address multiple spurious operation concems or other separation concerns.Consideration of component spurious actuation is not limited to the licensing basis criteria documented in UFSAR Appendix 10B, Section 5.9.4. As part of the transition to a NFPA 805 licensing basis, the criteria used in evaluating spurious actuation of components are those identified in NEI 00-01, Section 4, Identification and Treatment of Multiple Spurious Operations, which envelopes the plant licensing basis as discussed in the UFSAR. MSO component combinations were included in the assessments.

The number of potential spuriously operating valves in a line was not limited by number. The NSCA incorporates equipment identified during the review and the cable selection phase by providing an updated SSEL Report of the safe shutdown primary components and the Safe Shutdown Success Paths. In addition, the NSCA supports incorporation of secondary components into the database that were modeled as a result of the primary component selections.

A success path determination for all safe shutdown functions was performed.

Generally the path with the least amount of failures was utilized to demonstrate a success path for safe shutdown.

Support systems were reviewed in order to assess the impact on the systems being supported.

Primary and secondary shutdown methodologies have been developed.

The primary method employs the use of Emergency Condensers.

The secondary method employs the use of ERVs and ECCS equipment.

This approach provides for versatility by employing diverse equipment resulting in four potential shutdown success paths.Reference Document EIR 51-9133191-000, NSCA, Section 8.6 HNP RAI 3-14, NRC Request for Additional Information dated August 6, 2009 (ML092170715)

Technical Report on Identification

& Classification of the NMP-1 MSO Scenarios Using an Expert Panel -Review of New Scenarios, Rev.1 Closure of National Fire Protection Association 805 Transition Program Frequently Asked Question Number 08-0054 (ML110140183)

UFSAR Appendix lOB, Section 5.9.4 NMPI, April 2013 Page B-94 I I NMPI, April 2013 Page B-94 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.9 Fluid Density Effects 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 postfire safe shutdown capability.

This can be done systematically or via procedures such as Emergency Operating Procedures.

Applicability Comments Applicable Instrument tubing failure due to a fire is addressed in the NSCA (EIR 51-9133191, Section 1.14).Alignment Statement Aligns Alignment Basis Instrument sensing lines for level, pressure, flow, etc. that are exposed to a fire are considered to have the potential of causing erratic or unreliable signals or indication, unless a fire hazards analysis demonstrates that this failure is not credible.

Fire damage to instrument sensing lines can be as detrimental to the instruments as fire damage is to safe shutdown cables and components.

Even though the integrity of the tubing is expected to withstand the fire, the accuracy of the instrument may not be reflected correctly due to the heating of the fluid.The instrument sensing lines route locations are developed and inputted as design input to the analysis.

The input consisted of a list of instrument sensing lines (located outside Containment) including the fire areas and associated routing locations through the plant. This information was entered into the database as cables using fictitious cable numbers, including the route and endpoint identifications.

The analysis treated the tubing like cables and associated it with the instrument.

The sensing lines are subject to the same compliance issues and similar analytical techniques as safe shutdown cables. Sensing lines of instruments required for safe shutdown are included within the scope of a fire area assessment.

In this manner, the sensing lines are included for consideration along with cables when performing the fire area assessments.

If instruments were impacted by the fire, then alternate instruments, not impacted by the fire, would be relied upon for safe shutdown.Instrument sensing lines were reviewed for susceptibility to physical fire damage that may cause a loss of inventory.

Sensing lines for SSEL components are constructed of either stainless steel or carbon steel. Consequently, they are not susceptible to physical damage as the result of a postulated fire.Reference Document EIR 51-9133191, NSCA, Sections 5.4 and 8.4 HNP RAI 3-15, NRC Request for Additional Information dated August 6, 2009 (ML092170715)

NMPI, April 2013 Page B-95 I I NMP1, April 2013 Page B-95 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref 3.4.2 Methodology for Fire Area Assessment NEI 00-01 Guidance Refer to NEI 00-01 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 Applicability Comments Applicable None Alignment Statement Not Required Alignment Basis This paragraph provides introductory information and contains no specific requirements.

Discussion is provided in subsequent sub-sections.

Reference Document NMPI, April 2013 Page B-96 I!NMP1, April 2013 Page B-96 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref 3.4.2.1 Identify the Affected Equipment by Fire Area NEI 00-01 Guidance 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.Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis The information needed to support the Post Fire Safe Shutdown Analysis is maintained in a safe shutdown database which contains the safe shutdown systems, components, cables, and their associated fire area location.

This information is available in report format and can be sorted by fire area, system, train, component, cable, safe shutdown path, or various combinations of each. These reports are used to assess potential damage due to fire in each area of the plant. The database reports provide the same information identified on Attachment 5 of NEI 00-01.Reference Document EIR 51-9133191, NSCA, Sections 5.4 and 9.0 NMPI, April 2013 Page B-97 I I NMPI, April 2013 Page B-97 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref 3.4.2.2 Determine the Least Impacted Shutdown Path NEI 00-01 Guidance 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 frontiine systems that will require this power to support their operation.

Determine which components are required hot shutdown components and which components are important to SSD components using the guidance in Appendix H.Based on an assessment as described above, designate the required safe shutdown path(s) for the fire area. Classify the components on the required safe shutdown path necessary to perform the required safe shutdown functions as required safe shutdown components.

Identify all equipment not in the safe shutdown path whose spurious operation or mal-operation could affect the shutdown function.Criteria for classifying these components as required for hot shutdown or as important to SSD is contained in Appendix H. Include the affected 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.Applicability Applicable Comments None Alignment Statement Aligns Alignment Basis The Fire Area Compliance Assessments demonstrate the ability to achieve safe shutdown by ensuring at least one safe shutdown success path is available to accomplish the performance goals identified in NEI 00-01. NEI 00-01, Section 3 was used as guidance in performing the assessments.

The elements of the assessments performed for NMP1 reflect the NEI 00-01 guidance as discussed in the following.

Fire Area Assessments were performed on a Fire Area basis in order to ensure compliance in accordance with the requirements of NFPA 805.10 CFR 50, R. The safe shutdown database reports provide the potentially affected equipment and cables in each Fire Area, which were analyzed for impacts on safe shutdown success paths.The Safe Shutdown Equipment List (SSEL) contains equipment data such as the equipment type, description, safe shutdown path, drawing reference, fire area, fire zone, and room location.

Other equipment information would include normal position, hot shutdown position, cold shutdown position, failed air position, failed electrical position, high/low pressure interface concern, and spurious operation concern.I NMPI, April 2013 Page B-98 I I NMP1, April 2013 Page B-98 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review The route and location of all Appendix R cables (located by Fire Area) was used to generate the Fire Area Component Impact Reports, which identified affected systems and components on a Fire Area basis. The Fire Area Component Impact Reports were used as a means to determine the least impacted safe shutdown path for each fire area.The cable selection task involved identifying all cables associated with the control and operation of a safe shutdown component.

These cables were analyzed to determine the impact of fire induced cable failure on the selected equipment.

A circuit analysis was performed as part of the scope for selected cables/components, as required, in order to demonstrate that the cable is or is not required so that the analyzed component can be credited to perform its required function for the safe shutdown path.The Fire Area Analysis methodology identified fire-induced component and cable failures and spurious actuations, based on the cables and components present in the fire area of concern. All postulated cable and component failures were assessed and a resolution provided at the component level.Once the above was complete, the least impacted safe shutdown success path was identified so that mitigating strategies could be developed and documented in the Fire Area Assessment.

Reference Document EIR 51-9133191, NSCA, Sections 5.1, 5.2, 5.3, and 5.4 NMPI, April 2013 Page B-99 I I NMPI, April 2013 Page B-99 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref 3.4.2.3 Determine Safe Shutdown Equipment Impacts NEI 00-01 Guidance 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.Applicability Applicable Alignment Statement Aligns Comments None Alignment Basis The Safe Shutdown Equipment List was developed based on system requirements and other plant impacts. During the identification of the Safe Shutdown Equipment List for component cables, a circuit fault analysis for each component's cables was performed to determine the effects of a fire-induced hot short, open circuit and short-to-ground.

The circuits associated with the components operation and whose failure could affect the components operation was considered as required.

The fire area analysis assumed multiple fire-induced failures and multiple spurious actuations (MSOs), based on the cables and components present in the fire area of concern. The cable and component failures were evaluated and a resolution and disposition was provided for component and cable impacted in that fire area.In addition, spurious operating equipment concerns are addressed in the MSO Expert Panel Report," which consists of the multiple spurious operation review. The purpose of this review is to document the potential Multiple Spurious Operation combinations.

The results of this activity identifies equipment, whose fire-induced spurious operation could result in consequences that may be adverse to both the Fire PRA risk models and meeting the nuclear safety performance criteria of NFPA 805. The equipment identified in this task that could affect the Fire PRA will be integrated into the Fire PRA Equipment list.Reference Document EIR 51-9133191, NSCA, Sections 5.0, 8.0, and 9.0 ONS RAI 3-41 and RAI 3-43, NRC Request for Additional Information dated July 30, 2000 (ML102110394)

Technical Report on Identification

& Classification of the NMP-1 MSO Scenarios Using an Expert Panel -Review of New Scenarios, Rev.1 NMPI, April 2013 Page 6-100 I!NMP1, April 2013 Page B-100 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref NEI 00-01 Guidance 3.4.2.4 Develop a Compliance The available deterministic methods for mitigating the effects of circuit failures are summarized as follows (see Figure 1-1): Strategy or Disposition Required for Hot Shutdown Components: " Re-design the circuit or component to eliminate the concern. This option will require a revision to the post-fire safe shutdown analysis." Re-route the cable of concern. This option will require a revision to the post-fire safe shutdown analysis." Protect the cable in accordance with III.G.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." 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, LARs, Generic Letter 86-10 evaluation or fire protection design change evaluations with a licensing change process.Important to Safe Shutdown Components: " Any of the options provided for required for hot shutdown components." Perform and operator manual action in accordance with Appendix E." Address using fire modeling or a focused-scope fire PRA using the methods of Section 5 for MSO impacts. [Note: The use of fire modeling will require a review by the Expert Panel and the use of a focused-scope fire PRA will require a LAR.]Additional options are available for non-inerted containments as described in 10 CFR 50 Appendix R section IlI.G.2.d, e and f.Applicability Comments Applicable None Alignment Statement Aligns Alignment Basis The safe shutdown analysis provides a compliance strategy and various deterministic methods used for mitigating the effects of circuit failures.

Potential impacts to safe shutdown were addressed by using the path least impacted by the fire to assure at least one success path for safe shutdown.

This was accomplished by using a combination of the strategies listed in the guidance and taking credit for any existing features whenever possible.Circuit failures having the potential to adversely impact the shutdown process were identified as Open Items to be transitioned to the Fire Risk Evaluations.

Reference Document EIR 51-9133191, NSCA, Section 9.0!NMP1, April 2013 Page B-101 I Constellation Energy Nuclear Group Attachment B -NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref 3.4.2.5 Document the Compliance Strategy or Disposition NEI 00-01 Guidance Assign compliance strategy statements or codes to components or cables to identify 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.

Applicability Applicable Alignment Statement Aligns Comments Alignment Basis For the Safe Shutdown Analysis, success paths were developed and analyzed for each component impacted by a fire in the subject area using disposition codes that represent consistent standardized compliance statements.

The compliance statement reflects the credited fire protection features, analysis of credible cable failures, and serves as the basis for achieving safe shutdown conditions for the analyzed fire areas. The disposition codes (i.e., resolution of component hits) and associated statements were entered into the safe shutdown database.

The following is a sample of generic disposition codes used for the Fire Area Compliance Assessment:

  • Cable protected by rated fire barrier.* Failure of cable may result in loss of power/control of component.
  • Failure of cable may result in loss of indication or erroneous indication.
  • Component fails in desired SSD position/mode.
  • Series isolation valve(s) available and can be closed.* Capability to close valve is available from the MCR.* Component remains in desired SSD position.* Valve can spuriously open. Series isolation valve remains closed Reference Document EIR 51-9133191, NSCA, Section 9.0 NMPI, April 2013 Page B-102 I I NMP1, April 2013 Page B-102 I REVISIONS TO TRANSITION REPORT ATTACHMENT F FIRE-INDUCED MULTIPLE SPURIOUS OPERATIONS RESOLUTION Pages F-6 and F-7 with changes highlighted.

Constellation Energy Nuclear Group Attachment F -Fire-induced MS09 Resolution Results of Step 3: The results of the expert panel were included in Task 7.3.1 (NUREG/CR-6850 Task 2)and Task 7.4 (NUREG/CR-6850 Task 3) within the scope of the NMPI Fire PRA, and in Task 4.2.2, Table B-3 and Fire Area Analysis within the scope of the NMPI NSCA.Task 7.3.1 addressed spurious operations, including multiple spurious operations identified in the post-fire safe shutdown analysis, and those that resulted from the expert panel review.The results of the Fire PRA model update are included in NMPI Fire PRA Notebook,"Equipment Selection," which includes the following MSO related information: " Identification and disposition of equipment from the review of MSOs (Table D-1 of the "Equipment Selection" notebook);

and" Fire PRA equipment list, which includes MSO identified components and their associated basic events (Table G-1 of the "Equipment Selection" notebook).

The MSO combination components were also evaluated for inclusion into the NMPI NSCA. As necessary, components were added to the NSCA Equipment List and Logics, and the appropriate circuit analysis and cable routing were performed.

Step 4 -Evaluate for NFPA 805 Compliance The MSO combinations included in the NSCA should be evaluated with respect to compliance with the deterministic requirements of NFPA 805, as discussed in Section 4.2.3 of NFPA 805. For those situations in which the MSO combination does not meet the deterministic requirements of NFPA 805 (VFDR), the issue with the components and associated cables should be mitigated by other means (e.g., performance-based approach per Section 4.2.4 of NFPA 805, plant modification, etc.).The performance-based approach may include the use of feasible and reliable recovery actions. The use of recovery actions to demonstrate the availability of a success path for the nuclear safety performance criteria requires that the additional risk presented by the use of these recovery actions be evaluated (NFPA 805 Section 4.2.4).Results of Step 4: The MSO combination components of concern were evaluated as part of the NMPI NSCA and NPO analyses.

For NSCA cases where the pre-transition MSO combination components did not meet the deterministic compliance, the MSO combination components were added to the scope of the fire risk evaluations.

The process and results for Fire Risk Evaluations are summarized in Section 4.5 of the Transition Report.MSO scenarios impacting NPO POSs are associated with Key Safety Function (KSF)success paths in the NPO analysis.

A fire induced loss of these KSF success paths is addressed through the FAQ 07-0040 resolution process, wherein recommendations are provided to best manage fire risk in pinch point plant areas where KSFs may be impacted by a fire (EIR 51-9171174).

Step 5 -Document Results The results of the process should be documented.

The results should provide a detailed description of the MSO identification, analysis, disposition, and evaluation

!NMP1, April 2013 Page F-6 I Constellation Energy Nuclear Group Attachment F -Fire-induced MSOs Resolution results (e.g., references used to identify MSOs; the composition of the expert panel, the expert panel process, and the results of the expert panel process; disposition and evaluation results for each MSO, etc.). High level methodology utilized as part of the transition process should be included in the 10 CFR 50.48(c) License Amendment Request/Transition Report.Results of Step 5: The NMP1 Results are documented in: " "Resolution of Issues Related to Fire-Induced Circuit Failures, Technical Report on Identification

& Classification of the NMP1 MSO Scenarios using an Expert Panel"" NMP1 Fire PRA Notebook, N1 -ES-FOO1, "Equipment Selection (ES)"" NMP1 Fire PRA Notebook, N1-CS-F001, "Cable Selection, Detailed Circuit Analysis and Route Location (CS)"" NMP1 Fire PRA Notebook, N1-PRM-F001, "Plant Response Model"" NMP1 Fire Area Transition

-See Attachment C (NEI 04-02 Table B-3) of the Transition Report" EIR 51-9133191, NMP1 Nuclear Safety Capability Assessment (NSCA) Report" EIR 51-9137629, NMP1 Non-Power Operations KSF Equipment List" EIR 51-9171174, NMP1 NFPA 805 Transition-Non-Power Operations Component Pinch Point Analysis NMPI, April 2013 Page FT I REVISIONS TO TRANSITION REPORT ATTACHMENT G RECOVERY ACTIONS TRANSITION Pages G-1 through G-41 with changes highlighted.

Constellation Enerav Nuclear Grouo Attachment G -Recoverv Action Transition C.n.tellation...e...

Nuclear.....

Gru Attahmen G eoeyAtinTasto G. Recovery Actions Transition 36-31 Pages Attached NMPI, April 2013 Page G-1 I I NMPI, April 2013 Page G-1 I Constellation Energy Nuclear Group Attachment G -Recovery Action Transition In accordance with the guidance provided in NEI 04-02, FAQ 07-0030, Revision 5, and RG 1.205, the following methodology was used to determine recovery actions required for compliance (i.e., determining the population of post-transition recovery actions).

The methodology consisted of the following steps: " Step 1: Define the primary control station(s) and determine which pre-transition OMAs are taken at primary control station(s) (Activities that occur in the Main Control Room are not considered pre-transition OMAs). Activities that take place at primary control station(s) or in the Main Control Room are not recovery actions, by definition." Step 2: Determine the population of recovery actions that are required to resolve VFDRs (to meet the risk acceptance criteria or maintain a sufficient level of defense-in-depth)." Step 3: Evaluate the additional risk presented by the use of recovery actions required to demonstrate the availability of a success path." Step 4: Evaluate the feasibility of the recovery actions." Step 5: Evaluate the reliability of the recovery actions.An overview of these steps and the results of their implementation are provided below.Step 1 -Clearly define the primary control station(s) and determine which pre-transition OMAs are taken at primary control station(s)

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

Results of Step 1: Based on the definition provided in RG 1.205, and the additional guidance provided in FAQ 07-0030, the following locations are considered as taking place at the primary control station(s):

1. Remote Shutdown Panel 11 is located in Fire Area 7, Fire Zone T2B, Turbine Building Elevation 250'-0".2. Remote Shutdown Panel 12 is located in Fire Area 5, Fire Zone T4A, Turbine Building Elevation 277'-0".The remote shutdown panels were approved by the NRC in SER entitled "Subject Modifications and Alternate Safe Shutdown Capabilities to Comply with the Requirements of Appendix R", dated March 3, 1983.Table G-1 -"Recovery Actions and Activities Occurring at the Primary Control Station(s)" identifies the activities that occur at the primary control station(s).

Activities necessary to enable the primary control station(s) are also identified in Table G-1 as primary control station(s) activities.

These activities do not require the treatment of additional risk and are compliant with NFPA 805, Section 4.2.3.1.NMPI, April 2013 Page G-2 I I NMPI, April 2013 Page G-2 I Constellation Energy Nuclear Group Attachment G -Recovery Action Transition CoseltonEeg ulear Gru tahetG-Rcvr cinTasto Note that the Remote Shutdown Panels (RSPs) are primary control station(s) only for a fire in Fire Area 11 for which a MCR evacuation is credited.Step 2 -Determine the population of recovery actions that are required to resolve VFDRs (to meet the risk or defense-in-depth criteria)On a fire area basis, all VFDRs were identified in the NEI 04-02 Table B-3 (see Attachment C). Each VFDR not brought into compliance with the deterministic approach was evaluated using the performance-based approach of NFPA 805 Section 4.2.4. The performance-based evaluations resulted in the need for recovery actions to meet the risk acceptance criteria or maintain a sufficient level of defense-in-depth.

Results of Step 2: The FRE report provides the determination of recovery actions required to resolve VFDRs. These recovery actions are listed in Table G-1, "Recovery Actions and Activities Occurring at the Primary Control Station(s)." The actions contained in Table G-1 are identified on a fire area basis. Many of the same actions are repeated in different fire areas. To assist in understanding the various types of recovery actions contained in Table G-1, Table 4-1 has been created. Table 4-1 is a list of unique recovery actions only and does not include primary control station actions. It is important to note that not every component listed in the Components column of Table 4-1 is associated with every fire area listed in the Fire Areas of Concern column. The Fire Areas of Concern reflects the aggregate list of fire areas where the type of unique recovery action is credited to support shutdown.

It is also important to note that item 4---12 in Table 4-1 is a proposed modification identified in Attachment S, specifically Table S-1. The final set of recovery actions is provided in Table G-1 -Recovery Actions and Activities Occurring at the Primary Control Station(s).

NMPI, April 2013 Page G-3 I!NMPI, April 2013 Page G-3 I Constellation Enerav Nuclear GroUD Attachment G -Recoverv Action Transition Table 4-1: Unique Operator Recovery Actions No. Action Description Components Fire Areas of Concern 1 Operator determines vital LI-36-26, LI-36-28, LI-60-22C, LI-60-23C, 1,2, 5,10 parameters PI-201.2-94, PI-201.2-5, PI-36-25, PI-36-at RSP 27, TI-201-50B, TI-201-51B TI-201.2-521B, TI-201.2-522B, TI-32-02B, TI-32-03B, TI-32-04B, TI-32-05B 2 Operato lcally toproide W470-f3 4, 5, 6, 7, 0, 10, 11, 1 2, cooling MIto to 6hutdon4 43, 41 5,4I1A, 4gB, Gegl!1U.pS

17A, 2 OpVate locally to maintain RV-38 01, FVLV860-1 3, 5, 6, 7, 0, 10, 11, 12, conol domn rate 3811, iv38 01, 11 as 02, 1 11, 15,16A, 16B, E8n2 417A, 17B, 18,2920,2 4 Wirng rpaired QAd IV 39 01, IV 38 13, PMP 38 140, PMP 3 2 22,273,2044 conhgent opereatd 452 24 Valve locally throttled to VLV-60-11, VLV-60-12 4, 5,6,7,9,10,11, 12, control make-up to the 13, 14, 15,16A, 1613, Emergency Condensers 17A, 17B, 18,19,20,_________________21,22,23,24 36 Connect portable charger BAT-B1I, BAT-B312 5,6, 7,9, 10, 11 to charge Batteries__________

47- Manually isolate to prevent FCV-39-15, FCV-39-16, VLV-05-311, VLV- 5, 6, 9,10, 11,18, 22, inventory loss 05-32 23, 24 58 Vent air to close valve to IV-01-03, IV-01-04 5,6,7,9,10,11,18, prevent inventory loss 22, 23, 24 60 Vent air to open to IV-39-05, IV-39-06 5, 10 establish Emergency Condensers on failure to open 740 Verify tripped for load PB-BB131, PB-BB12, UPS-UPS162A, 5, 6, 7, 9,10, 11 shedding UPS-UPS162B, ,-UPS-UPS172A, UPS-UPS172B 448 Shut down locally for load PMP-79.1-01, PMP-79.1-07, PMP-79.1-20, 5,6,7, 9,10,11 shedding PMP-79.1-26 942 Locally operate to IV-39-07R, IV-39-08R, IV-39-09R, IV 10 establish decay heat IOR removal through Emergency Condensers 1042 Emergency Condenser LCV-60-17, LCV-60-18 10 Level Control Transfer switch to local 4411 Operate manually to VLV-93-13, VLV-93-16 5, 11,24 isolate Containment spray on spurious start 1245 Open NEW Disconnect to Recover Emergency Diesel Generator 5,7,9, 10, 11 load Emergency Diesel Generator to Dead bus (Currently a Damage Repair Procedure action)13461 Scram control rods by BV-113-3091, VLV-113-230 5, 7,10, 11 venting the scram air header 14 Locally operate the Fire BV-100-68, BV-100-69, PMP-100-02 4, 5, 6, 7, 9, 10, 11, 12, Water System using the 13,14,15, 16A, 16B, DFP to provide long term 17A, 17B, 18,19,20, Emergency Condenser 21, 22, 23, 24 makeup tank supply I I I NMPI, April 2013 Page G-4 I Constellation Enerav Nuclear Group Attachment G -Recovery Action Transition Coselto Eeg ula Gru AtahetG- eoeyAtinTasto Step 3: Evaluate the Additional Risk of the Use of Recovery Actions NFPA 805 Section 4.2.3.1 does not allow recovery actions when using the deterministic approach to meet the nuclear safety performance criteria.

However, the use of recovery actions is allowed by NFPA 805 using a risk informed, performance-based, approach, provided that the additional risk presented by the recovery actions is evaluated in accordance with NFPA 805 Section 4.2.4.Results of Step 3: The set of recovery actions that are necessary to demonstrate the availability of a success path for the nuclear safety performance criteria (see Table G-1) were evaluated for additional risk using the process described in NEI 04-02, FAQ 07-0030, Revision 5, and RG 1.205 and compared against the guidelines of RG 1.174 and RG 1.205. None of the recovery actions were found to have an adverse impact on the Fire PRA. The additional risk of recovery actions is provided in Attachment W.Step 4: Evaluate the Feasibility of Recovery Actions Recovery actions were evaluated against the feasibility criteria provided in NEI 04-02, FAQ 07-0030, Revision 5, and RG 1.205. Note that since actions taken at the primary control station are not recovery actions their feasibility is evaluated in accordance with procedures for validation of off normal procedures.

Results of Step 4: The HRA evaluated the feasibility of recovery actions modeled in the Fire PRA and used to resolve VFDRs identified in the B-3 Table. This includes recovery actions related to AC power, Emergency Diesel Generators, and long-term decay heat removal among others. Feasibility of these recovery actions were evaluated in the HRA against the criteria outlined in NEI 04-02, FAQ 07-0030 Revision 5, and RG 1.205, making extensive use of HEP quantifications.

Recovery actions that are required by the FRE but not addressed in the HRA were evaluated for feasibility using the NEI 04-02, FAQ 07-0030 Revision 5, and RG 1.205 criteria and documented in EIR 51-9156521 entitled, "Recovery Action Review for Nine Mile Point Nuclear Power Station Unit 1 Transition to NFPA 805." Since actions taken at primary control stations are not recovery actions, no independent feasibility evaluation is required.Results of the feasibility assessments in the HRA and in the EIR demonstrate that all credited NFPA 805 recovery actions are feasible.Implementation items resulting from the feasibility evaluation include: Modify, as needed, the following procedures for recovery actions being evaluated: " NI-SOP-21.1" N1-SOP-21.2 Operators will be trained and qualified on the revised procedures.

I NMPI, April 2013 Page G-5 I Constellation Energy Nuclear Group Attachment G -Recovery Action Transition These items are included as implementation items in Attachment S.Step 5: Evaluate the Reliability of Recovery Actions The evaluation of the reliability of recovery actions depends upon its characterization." The reliability of recovery actions that were modeled specifically in the Fire PRA were addressed using Fire PRA methods (i.e., HRA)." The reliability of recovery actions not modeled specifically in the Fire PRA are bounded by the treatment of additional risk associated with the applicable VFDR.In calculating the additional risk of the VFDR, the compliant case recovers the fire-induced failure(s) as if the variant condition no longer exists. The resulting delta risk between the variant and compliant condition bounds any additional risk for the recovery action even if that recovery action were modeled.Results of Step 5: The reliability of recovery actions that were modeled specifically in the Fire PRA were addressed using Fire PRA methods. The HRA addresses the reliability of these recovery actions, with consideration taken for various performance shaping factors, including cues and instrumentation, timing, procedures and training, complexity, workload pressure and stress, human-machine interface, environment, special equipment, specific fitness needs, as well as crew communications, staffing, and dynamics.

Accordingly, the HRA also evaluates recovery actions depending on whether they correspond or not to main control room abandonment situations.

Recovery actions that are required by the FRE but not addressed in the HRA are evaluated for reliability and documented in EIR 51-9156521 entitled, "Recovery Action Review for Nine Mile Point Nuclear Power Station Unit 1 Transition to NFPA 805." Since actions taken at primary control stations are not recovery actions, no independent reliability evaluation is required.

It should however be noted that a reliability evaluation documented in the HRA was made for those actions taken at PCSs that are credited and modeled in the Fire PRA.Results of the reliability assessments in the HRA and in EIR 51-9156521 demonstrate that all credited NFPA 805 recovery actions are reliable.NMPI, April 2013 Page G-6 I I NMPI, April 2013 Page G-6 I Constellation Enerav Nuclear GrouD Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Component Component Description Actions VFDR RA/PCS Area 1 LI-36-28 REACTOR VESSEL LEVEL Operator determines vital parameters VFDR-01-009 RA TI-201-50B DRYWELL TEMPERATURE from PNL-RSP 11. VFDR-01-011 1 LI-36-26 REACTOR VESSEL LEVEL Operator determines vital parameters VFDR-01-009 RA TI-201-51B DRYWELL TEMPERATURE from PNL-RSP12.

VFDR-01-011 2 LI-36-28 REACTOR VESSEL LEVEL Operator determines vital parameters VFDR-02-007 RA TI-201-50B DRYWELL TEMPERATURE from PNL-RSP11.

VFDR-02-008 TI-201.2-521 B TORUS TEMPERATURE VFDR-02-009 2 LI-36-26 REACTOR VESSEL LEVEL Operator determines vital parameters VFDR-02-007 RA TI-201-51B DRYWELL TEMPERATURE from PNL-RSP12.

VFDR-02-008 TI-201.2-522B TORUS TEMPERATURE VFDR-02-009 4 BV4O 63 44" AIR OPERATED BLOCKING 8V 70 63 is operatod locally to providc VFDR 04 006 RA VALVE BLC INLIE-TVALV 98oling Wator to the 69G PUMPS and a hWant a*nk far flip Q=C HXc 4 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-04-009 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the Emergency Condenser Makeup tanks.4 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-04-009 RA TURBINE FIRE PUMP run as needed to supply fire water to Emergency Condenser Makeup tanks.4 :RQV2-, Q 8" O.4 ^'PERATED FCV 38 09 is opo.atod locally to VFDR-04 8 RA FLO " CO T O.= .L VAV E cAn tQoe f&-- t-rough S -D C HX 3 8 1 362 to 2SIT-0UTDOW C=OLINGZ HEA rgulato RCS cool downi Fato.EXCH.ANGEcR 11 FLOW CON-TRO'-VAIU 4 AVK EPr DIAPRAG OPERATED FC3 TO 2O 10 is laoallytottlcl to VFDR-04-907 RA FLOWGCNTROL VALO E GOntE ol to Ewtr 1u 8G 12D HX 3p 122 Ie gHJTDWNl COIN HEAT C W egulato RsS corl decay Fateo EXCANER12FLODW 4 FGV-as i4 A" DIAPHR AGM OPERATED9 FCY 239 11 is opoatod locally to 1VFDR Q4 998 RA FLO COTRO VAVE onrol'10A flow throug SDC HX 38 120 to SHU TDOWNgh COO INGC HEA rogwlate RCS seel down rato.ECAGER 12 FLOW 4 VLV-60-1 1 MAKEUP VALVE TO LOOP 12 VLV-60-11I is locally throttled to control VFDR-04-004 RA EMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide a RCS heat sink for decay heat removal.I NMP1, April 2013 Page G-7 I Constellation Eneray Nuclear Group Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RA/PCS 4 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-04-003 RA EMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide a RCS heat sink for decay heat removal.5 BAT-B1 1 125 VOLT DC STATION Portable charger with a generator, VFDR-05-035 RA BATTERY 11 connected directly to Battery Board PB-BBI1 to charge Battery 811.5 BAT-B12 125 VOLT DC STATION Portable Charger with Generator VFDR-05-035 RA BATTERY NUMBER 12 connected directly to Battery Board PB-BB12 to charge Battery B12.5 BKR-(103/1-1)

DIESEL GEN 103 OUTPUT Operate disconnect switch locally. VFDR-05-040 RA R10321581 BREAKER 103/1-1 (R1032/581) to VFDR-05-043 PB-103 5 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-05-047 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the Emergency Condenser Makeup tanks.5 BV-1 13-3091 BLOCKING VALVE FOR AIR Operator unlocks and doses BV-1 13- VFDR-05-046 RA SUPPLY TO SCRAM AIR 3091. Operator removes the vent pipe SYSTEM cap, unlocks, and opens VLV-1 13-VLV-1 13-230 BALL VALVE -SCRAM VALVE 230 to vent the SCRAM air header to PILOT HEADER VENT insert the control rods.6 6-V38-04 SH WTDOWN COOLNG PUMP 12 BV 38 04 ic oporat"d !Wcally to alin VFDR-0-6044 RA SUCTION LOKIN 6LLV DC pump PMP 389 162 to #;e RCS to infitatok Ct e wd eo heat Fameval.6 RV70 !I," .AJR OpEp.RATED BLOC.KING BV 70 53 I6 ope-ated lca,3y to psodd.d VFDR 020 RA V.ALVE. RBCLCG INLET VALVE cooling ateF to the 6DC PUMPS and a TO SHUTDOWN~h COOL- IING heat cink for the SDC Hgs.-Y-S-T-rem 5 EG-EDG103 EMERGENCY GENERATOR

-Operate disconnect switch locally. VFDR-05-037 RA EMERGENCY DIESEL VFDR-05-038 GENERATOR UNIT 103 6 FGV 8- A2 DI.A-PHR-AG OPERA.TED FCV 38 10 ic operated locally to VFQR 06-021 RA FLOW CONTROL- I VAIVE contol fow throeugh GD HX 38 122 to SHU'-T-DOWAN COOLING HEA rogulate RCS cool downi Fate.EXCHANGE-R 12 RLON CN OLVALVEG NMPI, April 2013 Page G-8 IlNMP1, April 2013 Page G-8 I Constellation Enerav Nuclear GrouD Attachment G -Recoverv Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RA/PCS 5 FCV-39-15 EMERGENCY CONDENSER FCV-39-15 operated locally via the VFDR-05-003 RA STEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory loss CONTROL VALVE -FLOW flow path from the ECs 111 & 112 RCS CONTROL VALVE 11 STEAM return path drain line.LINE DRAIN 5 FCV-39-16 EMERGENCY CONDENSER FCV-39-16 operated locally via the VFDR-05-002 RA STEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory loss CONTROL VALVE -FLOW flow path from the ECs 121 & 122 RCS CONTROL VALVE 12 STEAM return path drain line.LINE DRAIN 5 IV-01-03 24" AIR OPERATED ( ANGLE) Air vented manually at MSIV IV-01-03 VFDR-05-009 RA ISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAIN STEAM ISOLATION VALVE 3 5 IV-01-04 24" AIR OPERATED ( ANGLE) Air vented manually at MSIV IV-01-04 VFDR-05-009 RA ISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAIN STEAM OUT ISOLATION VALVE 4 6 V 38- MOA.0TOIR 0OPE2RA.TE-D At PB 167, B KR (4 67!D3)62, the VFQR 06 925 R SHUTDOWN COOLING OUTLET valve ....Rn is ropairod and the "alv'INSIDE ISOLA.TION VALVE .p..atd to faclitate otabliehing a SDC suction flow path frcm the RCS to___________accosmplish docay heat Femev.al.IV-38 02 MOTOR OPERA.TED IV 38 02 i6 operatod lcally "ia the VFDR 95 4O4 RA SHUTDOWN COOLI'NG OUTLET hand ";,,cl to octablich a SDC Qucstion OUTSIDE ISOA.TION VALVE fle=ath fram the RCS te ar.....plch

________________________docay heat romoval.5 /2 2 REACTOR SHUTDOWN At 121 167, BKR (167!IG03)52, tho VFQR 95 025 R COOL'ING RETURN ISOLA.TION

'alve Ic repairod and the "al'e VA6VE 1 oporatod to facilitato establishing a 80C diechargo flow path to the RCS to_accemplich deray heat remoral.5 IV-39-05 EMERGENCY COOLING LOOP IV-39-05 is manually opened by VFDR-05-012 RA 11 CONDENSATE RETURN AIR venting air from the valve to establish a VFDR-05-013 OPERATED ISOLATION VALVE decay heat removal path using EC's VFDR-05-014 (GLOBE) 111 & 112.NMPI, April 2013 Page G-9 I I NMP1, April 2013 Page G-9 I Constellation Energy Nuclear Group Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Component Component Description Actions VFDR RA/PCS Area 5 IV-39-06 EMERGENCY COOLING LOOP IV-39-06 is manually opened by VFDR-05-012 RA 12 CONDENSATE RETURN AIR venting air from the valve to establish a VFDR-05-013 OPERATED ISOLATION VALVE decay heat removal path using EC's VFDR-05-014 (GLOBE) 121 & 122.5 PB-BB1 I 125VDC BATTERY BOARD 11 MG 167 motor BKR-MG167(BBI1/E03)

VFDR-05-028 RA verified tripped for battery load shedding to extend battery capability.

5 PB-BB12 125VDC BATTERY BOARD 12 MG 167 motor BKR-MG167(BB12/FO3)

VFDR-05-028 RA verified tripped for battery load shedding to extend battery capability.

5 PMP-79.1-01 EMERGENCY DIESEL PMP-79.1-01 shutdown locally for VFDR-05-028 RA GENERATOR 102 TURBO LUBE battery load shedding to extend battery OIL PUMP capability.

5 PMP-79.1-07 EMERGENCY DIESEL PMP-79.1-07 shutdown locally for VFDR-05-028 RA GENERATOR 102 CIRCULATING battery load shedding to extend battery LUBE OIL PUMP ( 6 GALLONS capability.

PER MINUTE)5 PMP-79.1-20 EMERGENCY DIESEL PMP-79.1-20 shutdown locally for VFDR-05-028 RA GENERATOR 103 TURBO LUBE battery load shedding to extend battery OIL PUMP capability.

5 PMP-79.1-26 EDG 103 CIRCULATING LUBE PMP-79.1-26 shutdown locally for VFDR-05-028 RA OIL PUMP (6 GPM) battery load shedding to extend battery capability.

5 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-05-047 RA TURBINE FIRE PUMP run as needed to supply fire water to Emergency Condenser Makeup tanks.5 LI-36-26 REACTOR VESSEL LEVEL Operator determines vital parameters VFDR-05-016 RA PI-36-27 REACTOR VESSEL PRESSURE from PNL-RSP12.

VFDR-05-017 5 UPS-UPS162A POWER SUPPLY -UPS 162A switches HDS-UPSI62A-VFDR-05-028 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load SUPPLY shedding to extend battery capability.

NMPI, April 2013 Page G-10 I!NMP1, April 2013 Page G-10 I Constellation Energy Nuclear Group Attachment G -Recovery Actions Transition Coselto nrg ula ru AtahetG RcvryAtosTasto Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RA/PCS 5 UPS-UPS162B POWER SUPPLY -UPS 162B switches HDS-UPS162B-VFDR-05-028 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load SUPPLY shedding to extend battery capability.

5 UPS-UPS172A POWER SUPPLY -UPS 172A switches HDS-UPS172A-VFDR-05-028 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load SUPPLY shedding to extend battery capability.

5 UPS-UPS172B POWER SUPPLY -UPS 172B switches HDS-UPS172B-VFDR-05-028 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load SUPPLY shedding to extend battery capability.

5 VLV-05-31 LOOP 11 VENT LINE MANUAL VLV-05-31 is operated locally via the VFDR-05-004 RA ISOLATION VALVE DOWN hand wheel to isolate an inventory loss STREAM OF 05-01R AND 05-11 flow path from the ECs 111 & 112 RCS steam supply vent line.5 VLV-05-32 LOOP 12 VENT LINE MANUAL VLV-05-32 is operated locally via the VFDR-05-004 RA ISOLATION VALVE hand wheel to isolate an inventory loss DOWNSTREAM OF 05-04R AND flow path from the EC's 121 & 122 05-12 RCS steam supply vent line.5 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-11 is locally throttled to control VFDR-05-018 RA EMERGENCY CONDENSERS makeup to EC's 121 & 122 to provide a RCS heat sink for decay heat removal.5 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-05-018 RA EMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide a RCS heat sink for decay heat removal.5 VLV-93-16 12" GATE VALVE -112 VLV-93-16 closed locally via the hand VFDR-05-022 RA CONTAINMENT SPRAY RAW wheel to isolate CTSRW flow to WATER PUMP DISCHARGE Containment Spray Header #12.VALVE 6 BAT-B11 125 VOLT DC STATION Portable charger with a generator VFDR-06-017 RA BATTERY 11 connected directly to Battery Board PB-BB11 to charge Battery B131.6 BAT-B12 125 VOLT DC STATION Portable Charger with Generator VFDR-06-017 RA BATTERY 12 connected directly to Battery Board PB-BB12 to charge Battery B12.I NMP1, April 2013 Page G-1 I I Constellation Energy Nuclear Group Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RA/PCS 6 V40--63 14" AIR OPERATED ALOCGKING BV 70 53 is oporatod locally to prow.de VFDR 06 0 RA wAt. .. ... ..... .ALVE 99.. In. Wat. .to the 60G pumps an. d a TO-'RW SHTDOWMN COOLING hcatl aink for t;A 2SDC HXA.6 FGV 38 99 8 DIHPRA.GM.4.

OPER-AXTED F.V 3110 is opeted locally to VFDRQ O 46- RA LOC ýI fWP threugh AGG HX 38 1 t6lo COOL ING HWEAT reg8t RCS cool don rato CONTRO' AI ý6BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-1 00-68 and BV-1 00-69 are locally VFDR-06-019 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the Emergency Condenser Makeup tanks.6 FCV-39-15 EMERGENCY CONDENSER FCV-39-15 operated locally via the VFDR-06-001 RA STEAM LINE DRAIN PRESSURE hand whieel to isolate an inventory loss CONTROL VALVE -FLOW flow path from the ECs 111 & 112 RCS CONTROL VALVE 11 STEAM return path drain line.LINE DRAIN 6 FCV-39-16 EMERGENCY CONDENSER FCV-39-16 operated locally via the VFDR-06-002 RA STEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory loss CONTROL VALVE -FLOW flow path from the ECs 121 & 122 RCS CONTROL VALVE 12 STEAM return path drain line.LINE DRAIN 6 IV-01-03 24- AIR OPERATED (ANGLE) Air vented manually at MSIV IV-01-03 VFDR-06-007 RA ISOLATION VALVE WITH P S hnweel to latprevent inventory loss.SOLENOID VALVES -MAIN STEAM ISOLATION VALVE 3 6 IV-01-04 24" AIR OPERATED (ANGLE) Air vented manually at MSIV IV-01-04 VFDR-06-008 RA ISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAIN STEAM OUT ISOLATION VALVE 4 6 IV 38 92 MQ r-O RAT-i W 38 02 ie oporated lecally '.ia the VFDR -0604 RA SH'JT-DOW.N COOLINGC OUTLET hQAd ":Oh..' to a SDC cuo':fon OUTSIDE ISOL.~TION VAOLVe fil'ath from t.he RCS ton ac.mplish I NMP1, April 2013 Page G-12 I Constellation Enerav Nuclear GroUD Attachment G -Recoverv Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Component Component Description Actions VFDR RAIPCS Area docay heat remoa4-.6 PB-BB11 125VDC BATTERY BOARD 11 MG 167 motor BKR-MG167(BB11/E03)

VFDR-06-017 RA verified tripped for battery load shedding to extend battery capability.

6 PB-BB12 125VDC BATTERY BOARD 12 MG 167 motor BKR-MG167(BB12/F03)

VFDR-06-017 RA verified tripped for battery load shedding to extend battery capability.

6 PMP-79.1-01 EMERGENCY DIESEL PMP-79.1-01 shutdown locally for VFDR-06-017 RA GENERATOR 102 TURBO LUBE battery load shedding to extend battery OIL PUMP capability.

6 PMP-79.1-07 EMERGENCY DIESEL PMP-79.1-07 shutdown locally for VFDR-06-017 RA GENERATOR 102 CIRCULATING battery load shedding to extend battery LUBE OIL PUMP (6 GALLONS capability.

PER MINUTE)6 PMP-79.1-20 EMERGENCY DIESEL PMP-79.1-20 shutdown locally for VFDR-06-017 RA GENERATOR 103 TURBO LUBE battery load shedding to extend battery OIL PUMP capability.

6 PMP-79.1-26 EDG 103 CIRCULATING LUBE PMP-79.1-26 shutdown locally for VFDR-06-017 RA OIL PUMP ( 6 GPM) battery load shedding to extend battery capability.

6 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-06-019 RA TURBINE FIRE PUMP run as needed to supply fire water to Emergency Condenser Makeup tanks.6 UPS-UPS162A POWER SUPPLY -UPS 162A switches HDS-UPS162A-VFDR-06-017 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load SUPPLY shedding to extend battery capability.

6 UPS-UPS162B POWER SUPPLY -UPS 162B switches HDS-UPS162B-VFDR-06-017 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load SUPPLY shedding to extend battery capability.

NMPI, AprIl 2013 Page G-13 I I NMPI, April 2013 Page G-13 I Constellation Enerav Nuclear Grouw Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR R.APCS 6 UPS-UPS172A POWER SUPPLY -UPS 172A switches HDS-UPS172A-VFDR-06-017 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load SUPPLY shedding to extend battery capability.

6 UPS-UPS172B POWER SUPPLY -UPS 172B switches HDS-UPS172B-VFDR-06-017 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load SUPPLY shedding to extend battery capability.

6 VLV-05-31 LOOP 11 VENT LINE MANUAL VLV-05-31 is operated locally via the VFDR-06-004 RA ISOLATION VALVE DOWN hand wheel to isolate an inventory loss VFDR-06-006 STREAM OF 05-01R AND 05-11 flow path from the ECs 111 & 112 RCS steam supply vent line.6 VLV-05-32 LOOP 12 VENT LINE MANUAL VLV-05-32 is operated locally via the VFDR-06-003 RA ISOLATION VALVE hand wheel to isolate an inventory loss VFDR-06-005 DOWNSTREAM OF 05-04R AND flow path from the ECs 121 & 122 RCS 05-12 steam supply vent line.6 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-06-011 RA EMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide a RCS heat sink for decay heat removal.6 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-06-011 RA EMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide a RCS heat sink for decay heat removal.7 BAT-B1 1 125 VOLT DC STATION Portable charger with a generator VFDR-07-012 RA BATTERY 11 connected directly to Battery Board PB-BB11 to charge Battery B131.7 BAT-B12 125 VOLT DC STATION Portable Charger with Generator VFDR-07-012 RA BATTERY 12 connected directly to Battery Board PB-BB12 to charge Battery B12.7 BV-113-3091 BLOCKING VALVE FOR AIR Operator unlocks and closes BV-113- VFDR-07-013 RA SUPPLY TO SCRAM AIR 3091. Operator removes the vent pipe SYSTEM cap, unlocks, and opens VLV-113-VLV-1 13-230 BALL VALVE -SCRAM VALVE 230 to vent the SCRAM air header to PILOT HEADER VENT insert the control rods.NMPI, April 2013 Page G-14 I I NMP1, April 2013 Page G-14 I Constellation Enerav Nuclear GrouD Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Component Component Description Actions VFDR RA/PCS Area 7 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-07-014 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the Emergency Condenser Makeup tanks.7 -V.70-63 44" AIR OERATED BLOCKIPNG I-V 70 53 is epwted loCally to VFDR-07 9-- RA~Is^,I~ ReCLC INLET. V'ALVE^= \ en .t... the ^ DC pU,,mpc and a"TO O STD Al COOL h101at11 -nk for tho CDC FzV- 3. 4. OF" OPPRAXTED FCV 38 410 *c op.-atod lc"ally to )ADR-07-04 RPA 414O6 COTRL .'VE control Aoof throug SD 6C HX 38 1322 to SHUTDOWN COOLIN HIN AT rg-uto RCS G oQol doa"-. rto.EXHAGR 1 2 FLON 7 IV-01-03 24" AIR OPERATED (ANGLE) Air vented manually at MSIV IV-01-03 VFDR-07-001 RA ISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAIN STEAM ISOLATION VALVE 3 7 IV-01-04 24" AIR OPERATED (ANGLE) Air vented manually at MSIV IV-01-04 VFDR-07-002 RA ISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAIN STEAM OUT ISOLATION VALVE 4 7 PB-BB11 125VDC BATTERY BOARD 11 MG 167 motor BKR-MG167(BB11/E03)

VFDR-07-012 RA verified tripped for battery load shedding to extend battery capability.

7 PB-BB12 125VDC BATTERY BOARD 12 MG 167 motor BKR-MG167(BB12/F03)

VFDR-07-012 RA verified tripped for battery load shedding to extend battery capability.

PMP-38-71.

SH1JTDOW.AN COO'ING PUMP 42 SDC pump PMP 33 1.2 io op.ratod at VRADR-07-0O RA All12A PB6 176, BKR (47 .'00eA)52 to R~A&AM& a 20C fle~ath to....amplih d..ay heat ..me.al.7 PMP-79.1-01 EMERGENCY DIESEL PMP-79.1-01 shutdown locally for VFDR-07-012 RA GENERATOR 102 TURBO LUBE battery load shedding to extend battery OIL PUMP capability.

NMPI, AprIl 2013 Page G-1 5 I I NMP1, April 2013 Page G-1 5 1 Constellation Enerav Nuclear GrouD Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Component Component Description Actions VFDR RA/PCS Area 7 PMP-79.1-07 EMERGENCY DIESEL PMP-79.1-07 shutdown locally for VFDR-07-012 RA GENERATOR 102 CIRCULATING battery load shedding to extend battery LUBE OIL PUMP (6 GALLONS capability.

PER MINUTE)7 PMP-79.1-20 EMERGENCY DIESEL PMP-79.1-20 shutdown locally for VFDR-07-012 RA GENERATOR 103 TURBO LUBE battery load shedding to extend battery OIL PUMP capability.

7 PMP-79.1-26 EDG 103 CIRCULATING LUBE PMP-79.1-26 shutdown locally for VFDR-07-012 RA OIL PUMP ( 6 GPM) battery load shedding to extend battery capability.

7 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-07-014 RA TURBINE FIRE PUMP run as needed to supply fire water to Emergency Condenser Makeup tanks.7 UPS-UPS162A POWER SUPPLY -UPS 162A switches HDS-UPS162A-VFDR-07-012 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load SUPPLY shedding to extend battery capability.

7 UPS-UPS162B POWER SUPPLY -UPS 162B switches HDS-UPS162B-VFDR-07-012 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load SUPPLY shedding to extend battery capability.

7 UPS-UPS172A POWER SUPPLY -UPS 172A switches HDS-UPS172A-VFDR-07-012 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load SUPPLY shedding to extend battery capability.

7 UPS-UPS172B POWER SUPPLY -UPS 172B switches HDS-UPS172B-VFDR-07-012 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load SUPPLY shedding to extend battery capability.

7 VLV-60-1 I MAKEUP VALVE TO LOOP 12 VLV-60-11 is locally throttled to control VFDR-07-007 RA EMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide a RCS heat sink for decay heat removal.7 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-07-007 RA EMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide a RCS heat sink for decay heat removal.NMPI, April2013 Page G-16 I I NMP1, April 2013 Page G-16 I Constellation Energy Nuclear Group Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RA/PCS 9 BAT-B11 125 VOLT DC STATION Portable charger with a generator VFDR-09-018 RA BATTERY 11 connected directly to Battery Board PB-BB11 to charge Battery B11.9 BAT-B12 125 VOLT DC STATION Portable Charger with Generator VFDR-09-018 RA BATTERY 12 connected directly to Battery Board PB-BB12 to charge Battery B12.9 BKR-(102/2-1)

DIESEL GEN 102 OUTPUT Operate disconnect switch locally to VFDR-09-019 RA R1022/571 BREAKER 2-1(R1022/571) to allow for recovery of diesel generator.

PB102 9 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-09-022 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the Emergency Condenser Makeup tanks.9 6V r-63 11" AIR 0-OPERATED BLOCKING 6V 7. 65 Is oprated casally to preode VFDR-O9 013 RA VALVEG RB-QCLC INILET VALVE seeling water to the 60C pumnps and a TSHTOWN COOL ING, hcAt Sink for the SDC HXc.S FCV4 29 an A" D'-A.PR.A.GM OPERATED FCV 38 00 is oporat-d Wocally to VIZ-309 015 RA FLW OTROL VALVE soto e floa'w -AA t-hre-ug h DC GHX 3 IS2 16% to SHU1TDOWN COOL--1INGC HET reoulato RCS cool- doa-.i rato.a FGV44 8" OPER-ATE-D FC-V 38 11 Is oporated locally to VFDR 09 04 4 RA FLOW C TROL VI V9ILV clWnl- Ilow through 6DG HX 3l 420 to SHUTODOWN COOL31ING HET rOgulato RCS coo! diNANi rato EX.CHA=G.Ar-ER 13 FWLMO^-r ------- VAI Vs 9 FCV-39-15 EMERGENCY CONDENSER FCV-39-15 operated locally via the VFDR-09-002 RA STEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory loss CONTROL VALVE -FLOW flow path from the ECs 111 & 112 RCS CONTROL VALVE 11 STEAM return path drain line.LINE DRAIN 9 FCV-39-16 EMERGENCY CONDENSER FCV-39-16 operated locally via the VFDR-09-003 RA STEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory loss CONTROL VALVE -FLOW flow path from the ECs 121 & 122 RCS CONTROL VALVE 12 STEAM return path drain line.LINE DRAINlNMPI, April 2013 Page G-17 I Constellation Energy Nuclear Group Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Component Component Description Actions VFDR RA/PCS Area 9 IV-01-04 24" AIR OPERATED (ANGLE) Air vented manually at MSIV IV-01-04 VFDR-09-008 RA ISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAIN STEAM OUT ISOLATION VALVE 4 P IV-38B1 12OOR OPERATED IV 39 02 us moporBKatMd loGally G Ba1E VFDR-09-021 RA SHTD I COO 1-ýING_ O-UTL E-T hand- boohe to cc-t-ablugch a 2SDC_ Acuction OU TS IDE. ISLTI CALVE fiedpath roem tho RS teo alompalh shdeday that erxn bmatyal.9 PB-BB1 1 125VDC BATTERY BOARD 11 MG 167 motor BKR-MG167(BB 12/E03) VFDR-09-018 RA verified tripped for battery load shedding to extend battery capability.

9 PB-BB12 125VDC BATTERY BOARD 12 MG 167 motor BKR-MG167(BB129FO3)

VFDR-09-018 RA verified tripped for battery load shedding to extend battery capability.

9 PMP-79.1-01 EMERGENCY DIESEL PMP-79.1-01 shutdown locally for VFDR-09-018 RA GENERATOR 102 TURBO LUBE battery load shedding to extend battery OIL PUMP capability.

9 PMP-79.1-07 EMERGENCY DIESEL PMP-79.1-07 shutdown locally for VFDR-09-018 RA GENERATOR 102 CIRCULATING battery load shedding to extend battery LUBE OIL PUMP (6 GALLONS capability.

PER MINUTE) ______________

_____9 PMP-79.1-20 EMERGENCY DIESEL PMP-79.1-20 shutdown locally for VFDR-09-018 RA GENERATOR 103 TURBO LUBE battery load shedding to extend battery OIL PUMP capability.

9 PMP-79.1-26 EDG 103 CIRCULATING LUBE PMP-79.1-26 shutdown locally for VFDR-09-018 RA OIL PUMP ( 6 GPM) battery load shedding to extend battery capability.

9 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-09-022 RA TURBINE FIRE PUMP run as needed to supply fire water to Emergency Condenser Makeup tanks.NMPI, April 2013 Page G-18 I i NMPI, April 2013 Page G-18 I Constellation Enemy Nuclear Group Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Component Component Description Actions VFDR RAIPCS Area 9 UPS-UPS162A POWER SUPPLY -UPS 162A switches HDS-UPS162A-VFDR-09-018 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load SUPPLY shedding to extend battery capability.

9 UPS-UPS162B POWER SUPPLY -UPS 162B switches HDS-UPS162B-VFDR-09-018 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load SUPPLY shedding to extend battery capability.

9 UPS-UPS172A POWER SUPPLY -UPS 172A switches HDS-UPS172A-VFDR-09-018 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load SUPPLY shedding to extend battery capability.

9 UPS-UPS172B POWER SUPPLY -UPS 172B switches HDS-UPS172B-VFDR-09-018 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load SUPPLY shedding to extend battery capability.

9 VLV-05-31 LOOP 11 VENT LINE MANUAL VLV-05-31 is operated locally via the VFDR-09-005 RA ISOLATION VALVE DOWN hand wheel to isolate an inventory loss VFDR-09-007 STREAM OF 05-01R AND 05-11 flow path from the ECs 111 & 112 RCS steam supply vent line.9 VLV-05-32 LOOP 12 VENT LINE MANUAL VLV-05-32 is operated locally via the VFDR-09-004 RA ISOLATION VALVE hand wheel to isolate an inventory loss VFDR-09-006 DOWNSTREAM OF 05-04R AND flow path from the ECs 121 & 122 RCS 05-12 steam supply vent line.9 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-11 is locally throttled to control VFDR-09-011 RA EMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide a RCS heat sink for decay heat removal.9 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-09-011 RA EMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide a RCS heat sink for decay heat removal.10 BAT-B11 125 VOLT DC STATION Portable charger with a generator VFDR-10-014 RA BATTERY 11 connected directly to Battery Board PB-BB11 to charge Battery B131.10 BAT-B12 125 VOLT DC STATION Portable Charger with Generator VFDR-10-014 RA BATTERY 12 connected directly to Battery Board PB-BB12 to charge Battery B12.I NMP1, April 2013 Page G-19 I Constellation Energy Nuclear Group Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RA/PCS 10 BKR-(102/2-1)

DIESEL GEN 102 OUTPUT Operate disconnect switch locally to VFDR-10-021 RA R1022/571 BREAKER 2-1(R1 022/571) TO allow for recovery of diesel generator.

PB102 10 BV-1 13-3091 BLOCKING VALVE FOR AIR Operator unlocks and closes BV-1 13- VFDR-10-028 RA SUPPLY TO SCRAM AIR 3091. Operator removes the vent pipe SYSTEM cap, unlocks, and opens VLV-1 13-VLV-1 13-230 BALL VALVE -SCRAM VALVE 230 to vent the SCRAM air header to PILOT HEADER VENT insert the control rods.10 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-10-029 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the Emergency Condenser Makeup tanks.40 BV40-53 14" AI R OPERATED BLOCKING BE 70 53 i6 0p-t GK.

to proyide VFDR-40-949 RA VAi\/E RQQCLCINLETVALVE cooling Wa!F 11o tho SDC pumps and a TOSHTDON COOLING heat WAI; for thA 20C HXc.SYSTEM 40 FGV 38 09 A" DIAPHWRAG" O1PERATED FCV 38 09 is oporated locsally to VFQR 10 020 R FLOWCONTOL VLVEcanal flow throu gh SDC HX 38 135 to RWI I-MAN QQ' IG Wrcgulate RCS cool devm Fate.EXCH A-NGER 11 FrLOW CONTIC3 VA '10 FCV-39-15 EMERGENCY CONDENSER FCV-39-15 operated locally via the VFDR-10-007 RA STEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory loss CONTROL VALVE -FLOW flow path from the ECs 111 & 112 RCS CONTROL VALVE 11 STEAM return path drain line.LINE DRAIN 10 FCV-39-16 EMERGENCY CONDENSER FCV-39-16 operated locally via the VFDR-10-008 RA STEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory loss CONTROL VALVE -FLOW flow path from the ECs 121 & 122 RCS CONTROL VALVE 12 STEAM return path drain line.LINE DRAIN 10 IV-01-03 24" AIR OPERATED (ANGLE) Air vented manually at MSIV IV-01-03 VFDR-10-009 RA ISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAIN STEAM ISOLATION VALVE 3 10 IV-01-04 24" AIR OPERATED (ANGLE) Air vented manually at MSIV IV-01-04 VFDR-10-010 RA ISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAIN STEAM OUT ISOLATION VALVE 4 I NMPI, April 2013 Page G-20 I Constellation Enerav Nuclear GrouD Attachment G -Recoverv Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RA/PCS 40 IV-38 01 MOT-OR OPERATE[)

AtP 67, 8KR (167!D03)52, the VFQR 10 0-11 RA SHUTDOWN CONOL C NG1 IS I OUTLET .9 ...ale w.ring ic .opairad and the .a...INSIDE ISOLATION VALVE- oporatod to farsilitato establiching a SDc cucfion flow path from the RCS to acoosmplich docay heat romov.al.40 IV-39-02 0MOTOR OPERATED IV 38 02 i patd lnally o ia thb VFDR-10-012 RA SH1-DOWNE COO- eINGtiT hand fAoml to eatablish a aOC cuotion OUTSE ISOLATION VALVE f(ecath fram the r RCoSa to atsinECh dewBy-heat rco&1 al.10 IV-39-05 EMERGENCY COOLING LOOP IV-39-05 is manually opened by VFDR-10-012 RA 11 CONDENSATE RETURN AIR venting air from the valve to establish a OPERATED ISOLATION VALVE ( decay heat removal path using ECs GLOBE) 11 & 112.10 IV-39-06 EMERGENCY COOLING LOOP IV-39-06 is manually opened by VFDR-10-012 RA 12 CONDENSATE RETURN AIR venting air from the valve to establish a OPERATED ISOLATION VALVE ( decay heat removal path using ECs GLOBE) 121 & 122.10 IV-39-07R MOTOR OPERATED LOOP 11 IV-39-07R is operated locally via the VFDR-10-012 RA STEAM OUTLET OUTSIDE hand wheel to establish a decay heat ISOLATION VALVE 112 removal flowpath using ECs 111 & 112.10 IV-39-08R MOTOR OPERATED LOOP 12 IV-39-08R is operated locally via the VFDR-10-012 RA STEAM OUTLET OUTSIDE hand wheel to establish a decay heat ISOLATION VALVE 122 removal flowpath using ECs 121 & 122.10 IV-39-09R MOTOR OPERATED LOOP 11 IV-39-09R is operated locally via the VFDR-10-012 RA STEAM OUTLET INSIDE hand wheel to establish a decay heat ISOLATION VALVE 111 removal flowpath using ECs 111 & 112.10 IV-39-IOR MOTOR OPERATED LOOP 12 IV-39-I1OR is operated locally via the VFDR-10-012 RA STEAM OUTLET INSIDE hand wheel to establish a decay heat ISOLATION VALVE 121 removal flowpath using ECs 121 & 122.10 LCV-60-17 EMERGENCY CONDENSER 111 Place EC 111/112 Level Control VFDR-10-012 RA-112 LEVEL CONTROL VALVE ( Transfer switch to Local and verify Auto LOOP 11 ) -AIR ACTUATED control by observing "A" on status FAIL OPEN panel at PNL-RSP11 to override false EC high level signal to support decay heat removal via EC's 111 & 112.I NMPI, April 2013 Page G-21 I Constellation Energy Nuclear Group Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Component Component Description Actions VFDR RA/PCS Area 10 LCV-60-18 EMERGENCY CONDENSER 121 Place EC 121/122 Level Control VFDR-10-012 RA-122 LEVEL CONTROL VALVE ( Transfer switch to Local and verify Auto LOOP 12 ) -AIR ACTUATED control by observing "A" on status FAIL OPEN panel at PNL-RSP12 to override false EC high level signal to support decay heat removal via EC's 121 & 122.10 LI-36-28 REACTOR VESSEL LEVEL Operator determines vital parameters VFDR-10-013 RA PI-36-25 REACTOR VESSEL PRESSURE from PNL-RSP11.

10 PB-BB11 125VDC BATTERY BOARD 11 MG 167 motor BKR-MG167(BB1I/E03)

VFDR-10-014 RA verified tripped for battery load shedding to extend battery capability.

10 PB-BB12 125VDC BATTERY BOARD 12 MG 167 motor BKR-MG167(BB12/F03)

VFDR-10-014 RA verified tripped for battery load shedding to extend battery capability.

40 PMP 38 4,9 SHU....TD

......N C LING PUMP SDC pIPA8_- 149 110'rng ropai:rd and VF.R...026R NUO2A oporatod locally at PB3 lOB BKR (I16B/0OOA)52 to cae c a 2SDCM fie~'ath to accomplich decay hoat 10 PMP-79.1-01 EMERGENCY DIESEL PMP-79.1-01 shutdown locally for VFDR-10-014 RA GENERATOR 102 TURBO LUBE battery load shedding to extend battery OIL PUMP capability.

10 PMP-79.1-07 EMERGENCY DIESEL PMP-79.1-07 shutdown locally for VFDR-10-014 RA GENERATOR 102 CIRCULATING battery load shedding to extend battery LUBE OIL PUMP (6 GALLONS capability.

PER MINUTE)10 PMP-79.1-20 EMERGENCY DIESEL PMP-79.1-20 shutdown locally for VFDR-10-014 RA GENERATOR 103 TURBO LUBE battery load shedding to extend battery OIL PUMP capability.

10 PMP-79.1-26 EDG 103 CIRCULATING LUBE PMP-79.1-26 shutdown locally for VFDR-10-014 RA OIL PUMP ( 6 GPM) battery load shedding to extend battery capability.

NMPI, April 2013 Page G-22 I I NMP1, April 2013 Page G-22 I Constellation Energy Nuclear Group Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RA/PCS 10 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-10-029 RA TURBINE FIRE PUMP run as needed to supply fire water to Emergency Condenser Makeup tanks.10 UPS-UPS162A POWER SUPPLY -UPS 162A switches HDS-UPS162A-VFDR-10-014 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load SUPPLY shedding to extend battery capability.

10 UPS-UPS162B POWER SUPPLY -UPS 162B switches HDS-UPS162B-VFDR-10-014 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load SUPPLY shedding to extend battery capability.

10 UPS-UPS172A POWER SUPPLY -UPS 172A switches HDS-UPS172A-VFDR-10-014 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load SUPPLY shedding to extend battery capability.

10 UPS-UPS172B POWER SUPPLY -UPS 172B switches HDS-UPS172B-VFDR-10-014 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load SUPPLY shedding to extend battery capability.

10 VLV-05-31 LOOP 11 VENT LINE MANUAL VLV-05-31 is operated locally via the VFDR-10-005 RA ISOLATION VALVE DOWN hand wheel to isolate an inventory loss VFDR-10-006 STREAM OF 05-01R AND 05-11 flow path from the ECs 111 & 112 RCS steam supply vent line.10 VLV-05-32 LOOP 12 VENT LINE MANUAL VLV-05-32 is operated locally via the VFDR-10-003 RA ISOLATION VALVE hand wheel to isolate an inventory loss VFDR-10-004 DOWNSTREAM OF 05-04R AND flow path from the ECs 121 & 122 RCS 05-12 steam supply vent line.10 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-10-017 RA EMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide a RCS heat sink for decay heat removal.10 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-10-017 RA EMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide a RCS heat sink for decay heat removal.11 BAT-B11 125 VOLT DC STATION Portable charger with a generator VFDR-1 1-020 RA BATTERY NUMBER 11 connected directly to Battery Board PB- VFDR-1 1-030 BB11 to charge Battery Bl1.NMPI, AprIl 2013 Page G-23 I I NMPI, April 2013 Page G-23 I Constellation Enerav Nuclear Grouo Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RAIPCS 11 BAT-B 12 125 VOLT DC STATION Portable Charger with Generator VFDR-1 1-020 RA BATTERY NUMBER 12 connected directly to Battery Board PB- VFDR-1 1-030 BB12 to charge Battery B12.11 BV-1 13-3091 BLOCKING VALVE FOR AIR Operator unlocks and doses BV-1 13- VFDR-1 1-036 RA SUPPLY TO SCRAM AIR 3091. Operator removes the vent pipe SYSTEM cap, unlocks, and opens VLV-113-VLV-113-230 BALL VALVE -SCRAM VALVE 230 to vent the SCRAM air header to PILOT HEADER VENT insert the control rods.11 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-11-037 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the Emergency Condenser Makeup tanks.14 BV-38-04 SuTDO, COOLING PUMP 12 2 8V 38 04 *c oporat.d locally to align VFDR- 4 --03 RA S'JCIONBLOKIN VAVE DC pump PMP 38 15 42 to the RCS to!.:ti SDC to pro-"ido dc::,"ay hcat 44- A-70 A- ^4-IR OPE^RATED LO=CKING BV 70 63 is epcmt-d locally to p.o.ido VFrDR 4- 04 RA VAL^ VE RoEQCL- INLET V'ALVE, , OlG .O.. +^. to the SOD pumps and a TOQ SHUlTDOWN COOLING hcat aink for tha SADC HXc.11 EMERGENCY EMERGENCY COOLING Place Emergency Cooling Isolation N/A PCS COOLING ISOLATION BYPASS SWITCH Bypass Switch in bypass to enable ISOLATION operation of IV-39-05, IV-39-07, and BYPASS IV-39-09 from PNL-RSP1 1.SWITCH 11 EMERGENCY EMERGENCY COOLING Place Emergency Cooling Isolation N/A PCS COOLING ISOLATION BYPASS SWITCH Bypass Switch in bypass to enable ISOLATION operation of IV-39-06, IV-39-08, and BYPASS IV-39-10 from PNL-RSP12.

SWITCH 4 FCA29 IQ DA-".W-PHRA

.':.-.OPET

-ED FOV 38 410 ic opoated locally to VFDR-14-036 RA FLWCNROL LA.L44 A acontra! flow 41th ra- ugh 6 DC HX 3 8 13- 2 to SHUl TDOWNI COOL ING HET rogulato RC QcGol damp rat.EXHAGR 12 FLOW COTO UAI ý11 FCV-39-15 EMERGENCY CONDENSER FCV-39-15 operated locally via the VFDR-1 1-009 RA STEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory loss CONTROL VALVE -FLOW flow path from the ECs 111 & 112 RCS CONTROL VALVE 11 STEAM return path drain line.LINE DRAIN I NMPI, April 2013 Page G-24 I Constellation Enerav Nuclear GrouD Attachment G -Recoverv Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RA/PCS 11 FCV-39-16 EMERGENCY CONDENSER FCV-39-16 operated locally via the VFDR-11-010 RA STEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory loss CONTROL VALVE -FLOW flow path from the ECs 121 & 122 RCS CONTROL VALVE 12 STEAM return path drain line.LINE DRAIN 11 IV-01-03 24" AIR OPERATED ( ANGLE) Air vented manually at MSIV IV-01-03 VFDR-1 1-003 RA ISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAIN STEAM ISOLATION VALVE 3 11 IV-01-04 24" AIR OPERATED (ANGLE) Air vented manually at MSlV IV-01-04 VFDR-11-004 RA ISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAIN STEAM OUT ISOLATION VALVE 4 14- IV,3849 MOTOR OPERATED At PB 167, BKR (1674D03)52, the QR RA SHUDON OOLING OUTL ET valve YAFRig is ropairod and the valve INID SOATION VAL VE oporatcd to facilitate establishing a SDC 13cQuc-,..tien fo' path frome the RCS to accompslih doray heat rFemea-.14- ,V-38 02 MOTOR OPERATED IV 38 02 ic epcrated lacally "ia the VFDR 11 035 RA SHUTDOWN COOLING OUTLET. hand -Ae' l to " stabr hlic-h a SD Cri suc t 6ion OUIT-SIDE ISA-TION VAL VE f .wpath from the RCS to ah C Splith 1CNENAERTUNAR dechay heat reMoaapa 14 IV 38 i3 REACT-OR SHUTDOWN At PB 167,13141 (167!G03)52, the VFDR-4-4-O29 RA OONRETURN ISIOAATIO N Nalve ALVEg i trpaiud and the valve VAI Vr= operated to facilitate ectablishing a 8DC diochargo flow path to the RCS to GacOBEamplish dersay heat rcmaval.11 IV-39-05 EMERGENCY COOLING LOOP IV-39-05 operated from PNL-RSP1 1 to N/A PCs 11 CONDENSATE RETURN AIR establish a decay heat removal path OPERATED ISOLATION VALVE ( through ECs 111 & 112.GLOBE )11 IV-39-06 EMERGENCY COOLING LOOP IV-39-06 operated from PNL-RSP12 to N/A PCS 12 CONDENSATE RETURN AIR establish a decay heat removal path OPERATED ISOLATION VALVE ( through ECs 121 & 122.___________GLOBE) 11 IV-39-07R MOTOR OPERATED LOOP 11 IV-39-07R eperated from PNL-RSP1 1 N/A PCs STEAM OUTLET OUTSIDE to establish a decay heat removal path ISOLATION VALVE 112 through ECs 111 & 112.I NMPI, April 2013 Page G-25 I Constellation Energy Nuclear Group Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RA/PCS 11 IV-39-08R MOTOR OPERATED LOOP 12 IV-39-08R operated from PNL-RSP12 N/A PCS STEAM OUTLET OUTSIDE to establish a decay heat removal path ISOLATION VALVE 122 through ECs 121 & 122.11 IV-39-09R MOTOR OPERATED LOOP 11 IV-39-09R operated from PNL-RSP1 1 N/A PCS STEAM OUTLET INSIDE to establish a decay heat removal path ISOLATION VALVE 111 through ECs 111 & 112.11 IV-39-1OR MOTOR OPERATED LOOP 12 IV-39-1OR operated from PNL-RSP12 N/A PCS STEAM OUTLET INSIDE to establish a decay heat removal path ISOLATION VALVE 121 through ECs 121 & 122.11 LCV-60-17 EMERGENCY CONDENSER 111 Place EC 111/112 Level Control N/A PCS-112 LEVEL CONTROL VALVE ( Transfer switch to Local and verify Auto LOOP 11 ) -AIR ACTUATED control by observing "A" on status FAIL OPEN panel at PNL-RSP11.

11 LCV-60-18 EMERGENCY CONDENSER 121 Place EC 121/122 Level Control N/A PCS-122 LEVEL CONTROL VALVE ( Transfer switch to Local and verify Auto LOOP 12 ) -AIR ACTUATED control by observing "A" on status FAIL OPEN panel at PNL-RSP12.

11 LI-36-26 REACTOR VESSEL LEVEL Operator determines vital parameters N/A PCS LI-60-23C EMERGENCY CONDENSER 121 from PNL-RSP12.

& 122 PI-201.2-94 DRYWELL PRESSURE PI-36-27 REACTOR VESSEL PRESSURE TI-201-51B DRYWELL TEMPERATURE TI-201.2-522B TORUS TEMPERATURE TI-32-04B REACTOR COOLANT TEMPERATURE TI-32-05B REACTOR COOLANT TEMPERATURE 11 MG-MG131 MG SET 131 Place MG Set #131 switch in the TRIP N/A PCS position and confirm CONTROL RODS IN white light lit on PNL-RSP11.

11 MG-MG141 MG SET 141 Place MG Set #141 switch in the TRIP N/A PCS position and confirm CONTROL RODS IN white light lit on PNL-RSP12.

NMPI, April 2013 Page G-26 I!NMPI, April 2013 Page G-26 I Constellation Eneray Nuclear Group Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RAIPCS 11 PB-BB11 125VDC BATTERY BOARD 11 MG 167 motor BKR-MG167(BB11/E03)

VFDR-11-020 RA verified tripped for battery load VFDR-1 1-030 shedding to extend battery capability.

11 PB-BB12 125VDC BATTERY BOARD 12 MG 167 motor BKR-MG167(BB12/F03)

VFDR-11-020 RA verified tripped for battery load VFDR-1 1-030 shedding to extend battery capability.

44 PMP-38-165 SHUJTDOWNA!.h COOL3NG PUMP 12 SDC pump PMP 38 162 Is opo.ated at V-FDR4 042 RA ISU 122 P29 476, BKfl (172E'00eA)92 to astabl-ah 2 29C fleovath to_______________________acwmpliish docay heat Faemoval.11 PMP-79.1-01 EMERGENCY DIESEL PMP-79.1-01 shutdown locally for VFDR-1 1-020 RA GENERATOR 102 TURBO LUBE battery load shedding to extend battery VFDR-1 1-030 OIL PUMP capability.

11 PMP-79.1-07 EMERGENCY DIESEL PMP-79.1-07 shutdown locally for VFDR-1 1-020 RA GENERATOR 102 CIRCULATING battery load shedding to extend battery VFDR-1 1-030 LUBE OIL PUMP (6 GALLONS capability.

PER MINUTE)11 PMP-79.1-20 EMERGENCY DIESEL PMP-79.1-20 shutdown locally for VFDR-11-020 RA GENERATOR 103 TURBO LUBE battery load shedding to extend battery VFDR-1 1-030 OIL PUMP capability.

11 PMP-79.1-26 EDG 103 CIRCULATING LUBE PMP-79.1-26 shutdown locally for VFDR-11-020 RA OIL PUMP (6 GPM) battery load shedding to extend battery VFDR-1 1-030 capability.

11 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-1 1-037 RA TURBINE FIRE PUMP run as needed to supply fire water to Emergency Condenser Makeup tanks.11 UPS-UPS162A POWER SUPPLY- UPS 162A switches HDS-UPS162A-VFDR-11-020 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load VFDR-1 1-030 SUPPLY shedding to extend battery capability.

11 UPS-UPS162B POWER SUPPLY -UPS 162B switches HDS-UPS162B-VFDR-11-020 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load VFDR-1 1-030 SUPPLY shedding to extend battery capability.

I NMPI, April 2013 Page G-27 I Constellation Enerav Nuclear Groue Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Component Component Description Actions VFDR RAIPCS Area 11 UPS-UPS172A POWER SUPPLY- UPS 172A switches HDS-UPS172A-VFDR-11-020 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load VFDR-1 1-030 SUPPLY shedding to extend battery capability.

11 UPS-UPS172B POWER SUPPLY -UPS 172B switches HDS-UPS172B-VFDR-11-020 RA UNINTERRUPTABLE POWER OS, BS, DC, IS opened for battery load VFDR-1 1-030 SUPPLY shedding to extend battery capability.

11 VLV-05-31 LOOP 11 VENT LINE MANUAL VLV-05-31 is operated locally via the VFDR-1 1-005 RA ISOLATION VALVE DOWN hand wheel to isolate an inventory loss VFDR-1 1-007 STREAM OF 05-01R AND 05-11 flow path from the ECs 111 & 112 RCS steam supply vent line.11 VLV-05-32 LOOP 12 VENT LINE MANUAL VLV-05-32 is operated locally via the VFDR-1 1-006 RA ISOLATION VALVE hand wheel to isolate an inventory loss VFDR-1 1-008 DOWNSTREAM OF 05-04R AND flow path from the ECs 121 & 122 RCS 05-12 steam supply vent line.11 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-11 is locally throttled to control VFDR-11-015 RA EMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide a RCS heat sink for decay heat removal.11 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-11-015 RA EMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide a RCS heat sink for decay heat removal.11 VLV-93-13 12" GATE VALVE -121 VLV-93-13 dosed locally via the hand VFDR-11-018 RA CONTAINMENT SPRAY RAW wheel to isolate CTSRW flow to WATER PUMP DISCHARGE Containment Spray Header #11.VALVE 11 VLV-93-16 12" GATE VALVE -112 VLV-93-16 closed locally via the hand VFDR-11-017 RA CONTAINMENT SPRAY RAW wheel to isolate CTSRW flow to WATER PUMP DISCHARGE Containment Spray Header #12.VALVE 12 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-12-009 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the Emergency Condenser Makeup tanks.412 BV-A) 63 4A11141I OPE-6-R-ATED- -RLOCGK!NG BY70 53 is oporatod loall opo.d VFDR.42 006 RA V.ALV4 R ,CL INCLET VAIVE 0 .at.. to the SDC pumps and a To SHUTD O he at sonk for tho SDG Wgs SYST-rm I NMPI, April 2013 Page G-28 I Constellation Enerav Nuclear GrouD Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RAIPCS 42 FGV A" DIAPHRAGM OPERATED FCV 38 00 is op"rat.d locally to Vr-R-4DI042 RA FLOW ~ ~ ~ -eA&- COTOfiLEc ntrolfowthrugh 6DC HX 38 125 to HuTD-Oil COOLING HET r^gulate RCS cool do.- Fate.EXCHANG R 11 FL ON 42 PMPA-102A DIELAPHRA OPERATED FDV 38 410 is mporanud loally to a nFDR-42-O09 RA FLORBONITROL VALU E cntal flow thosuppgh 6C rX 38 132 to 12SHV60-KUTWPNAL COO' ING 2 LV Eeguk-t1 RiS soel dly Frote.EXHAGR 12 FLOW RCShet in fr ecy ea rmoal 42 GVLV-s 80- " DMAKPHRVALV PERATED FCV 38 11 is ope1ated lcally to VFDR-42-3O-RA FLOWGCNTR VALOVE9S aoentro foEsw thrugh 69G HX 38 p2d to H1TDOWRN COOLING HEAT rOgulato RCS cool do" Fateo EXCHANIGER 41 FLOW a6U4 _-= 4AI ýra 12 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-12-009 RA TURBINE FIRE PUMP run as needed to supply fire water to_________Emergency Condenser Makeup tanks.12 VLV-60-1 1 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-12-004 RA EMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide a RCS heat sink for decay heat removal.12 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-12-003 RA EMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide a RCS heat sink for decay heat removal.44 6V zo 63 :11" AR OPERATED SLO-KINGQ BV 70 53 we oporated locally to pmoýAde VFDR 13 008 PA VALE RCLCINL-ET 9ALE colig at8F to tho SDC pumps and a TO SHUTDW COLNG hat snok for tho 213C HXc.43 F= 00 A* DLAPHIlA.GM OPERATED FCV 38 00 Is oporatod Woally to VFDR 42-0 PA FLOW COT OL VALE QWontrol flow through SDC HX 38 1356 to HU1TDOWNN COO ING HEAT reogulto RC 2caol dmop Fate.44 FGV as io a" DL "'PA.M O1PERATD FC 39 10is oporatod locally to FR308P FLO CONTROL VAL VE contAfrol f lowA 10hreough rsDC HX 3 8 13- 2 to-SHU1_TDOWN COOLING. HEAT rOgulato RCS Gaol doein rate-EXCHWANGQER 12 FLOWM VC~II'ALV I NMP1, April 2013 Page G-29 I Constellation Energy Nuclear Group Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RAJPCS 43 FGVas -44 8" DIA-P-A.GM PA F "V 38 14 is operated W ,ally to VFWDR--42-007 RA FLW CONTROL VALVE ont-erol flvow tlhroaugh SDC HX 38 120 to C =U IAIT Q 1 ING EAI 13 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-13-011 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the Emergency Condenser Makeup tanks.13 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-13-011 RA TURBINE FIRE PUMP run as needed to supply fire water to Emergency Condenser Makeup tanks.13 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-1 3-004 RA EMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide a RCS heat sink for decay heat removal.13 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-13-003 RA EMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide a RCS heat sink for decay heat removal.44 8VA)63 4"A-IR O ERATEDT BLO-KING.

8V 70 63 is opetated loWally to p-.ide VF'DR 4 00 RA VAL VE RBCLCG INLET VALVE cooling at8F to the 6DC pumps and a TO SHUTDOW COO.ING hAt. Sink for the 2DC XC..44 FGV 38 09 A" DIAPIHRArGM OPE0.RATED FCV 32-9 09 is epcated lcally to VFDR 44-006 RA FLO CO -NT-ROL VALVEcoto owtruhDCH3815o SWI-IODAW OG' NG WET rc8ulate RCS_ cool- doemn rat.EXCHAG-ER 41 FLOAW 44 FGV a8s 4 8" DIAOPHRAG!

M OPEsR-A.TED FCV 39 10 is operated locally to VFDR 4 RA FLOWhCOITROL VALVE con tro fle4A throUh SDG HX 38 132 to SHUTIDOWNA COOLIN HET rogulato RCS- cool doevip rate EXCANCR 12 FLOWMA COriTRO' 'AI ýra 44 FCA4 2-44 8" DMI.A.HR-AG1M OPER614ATED FCV 38 11 is opc~ated locally to VFDR 144008 RA FLO-W CONTROL VALVE coent-rol flow~ t-hrough SDC HX 38 1-20 to SHU1TDOWNAgb COOLI0 G HWET rgulate RCS G-eol dkwm rael.ECHANGRO 13I FLO NMPI, AprIl 2013 Page G-30 I I NMP1, April 2013 Page G-30 I Constellation Enerav Nuclear GrouD Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RA/PCS 14 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-14-009 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the Emergency Condenser Makeup tanks.14 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-14-009 RA TURBINE FIRE PUMP run as needed to supply fire water to Emergency Condenser Makeup tanks.14 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-14-004 RA EMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide a RCS heat sink for decay heat removal.14 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-14-003 RA EMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide a RCS heat sink for decay heat removal.4-6 BV-74-63 44" AIR OPERATED AL OCKIN_ 8Y 70 63 is operot.d lesal!y to pro.de VFDR- 46 00 RA wAV, R, CL- INLT VALVE c..ling "r to tho 0_C puFmpAS and a TO SHUI-TDO)-WN COO0L0 IING heat sinkI forf tho 2DC HXr.4- FCV4 9 8n " DL2.2 3PHRANGAM OPERPAT-ED FCV 38 09 is -pe-ated ecsally to VF OR -- RA BLO9 ONUTROL VALVES opentol pwerodicallgh 6 HX 38 136 to HUTDOWNl COO ING HET regubtge RC 9-sool denser rata.EXHNER 411 FlO" 41 PGVMP-100-DI APHRAGE OPERATID FV 38 P 10 is mpratnd locally te VFDR-15-9009 RA FO C TROL VALVE ntro I lneeod tough pDC l X 38 122 to 9hrgnc Condnse Makeu taks SHUDOW COLIG-EAT rogulato RCSG easol de;.o rato.EXHNER 12 FL ON CONR~O' VAI *9 44 FGV 39 -4 8" DLA.PHR-A.G1Q OPER-ATED FCY 38 is oporatod locally to VFDR 16 908 RA FLOWA GCONT-ROL VALVE control- flow through 69C H4X 38 1-20 le WHUTDO l OI2NQ HEA roguilat RCA cool dawn Fate, EXCHANGER 13 FLOW0 h'AI 49~15 BV-100-68 LOOP I111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-15-009 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the__________Emergency Condenser Makeup tanks.15 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-15-009 RA TURBINE FIRE PUMP run as needed to supply fire water to_________

___________

______________________Emergency Condenser Makeup tanks.NMPI, April 2013 Page G-31 I I NMP1, April 2013 Page G-31 I Constellation Energy Nuclear Group Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RA/PCS 15 VLV-80-1 1 MAKEUP VALVE TO LOOP 12 VLV-60-1 I is locally throttled to control VFDR-15-004 RA EMERGENCY CONDENSERS makeup to EC's 121 & 122 to provide a RCS heat sink for decay heat removal.15 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-15-003 RA EMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide a RCS heat sink for decay heat removal.46A RV A- ^2 44".^IR OPERATED BLOCKINGQ BV 70 63 is oporated l.cally to I;om.da VF-DR-4A-00 RA^J 0 _RG 1R-CC IN L1T, T' L V^ 996^0A9 ...c to the S6C pUMPS and a TO SHUTON COLNG hat rinkt for #;A ROC HXc 46A FV 80 R" 1 DIAPH RAG OPERA.TED FCV 38-09 i6 oprated locally to VF;-R- 0-l60 RA FLOW-A CONTROL VALVE control- flow throweugh 69C HX 38 1326 to SHTDW cool ING Hr.-AT Fogulate RCS_ caoe I M-NA. rato EXHNER 11 FLO0W COTR AI 'IE 446A F;QA -344 9_2 D!IA.1HR-AG O. -PER-ATED FCV 38 114 is oporated locally to VFDMR ISA 006 RA FLW OTROL VAL VE control flowA thro-ugh SOC HX 38 1 20 to AHlDW COO, ING HEAT regulate =CS cooel dao40i Fate, 46A IV 38 02 AMQQ- R0- PERATM WV 28 02 D~ operated locally ý.ia the VFDR 46A 008 RA SHUTDO30WN COOL INGQ OU1TZIT hand whaal to octablich a Q SOC uotio OUTSDE SOL.TIN VLVE fewal; katfom the RCS to accamplich 16A BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-1100-68 and BV-1 00-69 are locally VFDR-16A-009 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the_________Emergency Condenser Makeup tanks.16A PMP-.100-02 DIESEL DRIVEN VERTICAL PMP-1 00-02 is manually started and VFDR-16A.009 RA TURBINE FIRE PUMP run as needed to supply fire water to______________________

______________________Emergency Condenser Makeup tanks.16A VLV-60-1 1 MAKEUP VALVE TO LOOP 12 VLV-60-1 I is locally throttled to control VFDR-1 6A-004 RA EMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide a RCS heat sink for decay heat removal.NMPI, April 2013 Page G-32 I I NMP1, April 2013 Page G-32 I Constellation Enerav Nuclear GrouD Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Component Component Description Actions VFDR RAJPCS Area 16A VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-16A-003 RA EMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide a RCS heat sink for decay heat removal.46B BV-7-63 44" AIR ^OPERATED B6LOKING 6V 70 53 is ep.rat'd Iocally to provde VFDR-41"16B-RA ALER2CLC INLETVW 98HA cooIng Wtr the 60G pumps and a TO HTDOW OLN'G' hcat a-Alk for the SDQC WXA.4468 FGV-3& 40 AV DL'-.PHRAOM-4 OPER14ATED FCY 38 1 0 is cperated locsally to VFR4OB.lg496 RA..O OTROL VAL.VE can.. ..fl0w o ugh .DG 1. X 3.8 41 32 2 to SHTONI coOlIW NG E rogulatS RCS G-0-0l dW_.A Fato EXCHANGER412 FLOW 16B BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-16B-007 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the Emergency Condenser Makeup tanks.16B PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-16B-007 RA TURBINE FIRE PUMP run as needed to supply fire water to Emergency Condenser Makeup tanks.16B VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-16B-004 RA EMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide a RCS heat sink for decay heat removal.16B VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-16B-003 RA EMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide a RCS heat sink for decay heat removal.47A 63 11" AIR O ^ERATED AL OCVKNG 8V 70 63 Is op ..-atcd locally to VZDR47A"007 RA I- PRBCLC INLET

Vtor to #is 6DC pumps and a TO SHUll-T.DOWAN COO-INIG hcot cink for thA 21DC 1-14c S-YST-EM 47A FGV43-09 8-OPERATED FCV 328 00 is operated loWally to VFDR 47-A 00 RA F6LOWNTO VALVE flo, tR, ugh 60C HX 38 136 te HUDW C L HE--AT rOgulato RCS, eol ddo rat4.EXCHAANGER 11 FLOW NMPI, April 2013 Page G-33 I I NMPI, April 2013 Page G-33 I Constellation Enerav Nuclear GrouD Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RA/PCS 7A FGV-138 LOOP" 11 2 A ND OPERATED FCOV 11 BV 10 0 c6 opond ted lBcally to VFDR-17A-006 RA FLO CO2NTARL VALVES GnAI flo W through SDC HX 3r 12i to SWUTDOWN01 COOEING H eregulate RCS odenso r Fakpta.EXHNER 13 FLOW9 41A 04-100-02 DIO R 0VEN VRATEID IV 3P 02 is mpoanted lally saea VFDR-47A-O09 RA;HTON COOL1ING OUTLEfT hand wheel te es&t-ablish a 2-D-C Al 'GiQn OUTSIDE 1I01 ATION VALV flcathl *aom the RCS to accmplich decay hoo-t rcm a.17A BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-1 00-68 and BV-1 00-69 are locally VFDR-1 7A-009 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the Emergency Condenser Makeup tanks.17A PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-17A-009 RA TURBINE FIRE PUMP run as needed to supply fire water to)__________Emergency Condenser Makeup tanks.17A VLV-60-1 1 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-17A-004 RA EMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide a RCS heat sink for decay heat removal.17A VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-17A-003 RA EMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide a RCS heat sink for decay heat removal.1-TB AV70 62 14" AIR OPERIATED0 W1LOCKING RV 7-0 53 is operated locally to pro'~de vFDR-17-8 00O RA VALVE RBCLC INLET VA'VE cooling Vtcr 1to the SDC pumps and a S-Y-STEM TOV COOL ttING ha ~kfrtoSCHs 47-8 F-GV as&40 8 DI.APHWR-AGM O. -PEERATED FCV 38 10 Is oporatod locally to VFDR-4-78-0()5 RA FLW OTROL VALVE contWrolI flow.A t-hro-ugh 8DC H4X 38 1322 to HT OWN COOLING HEA regulato RCS cool down Fate.EXCW.N--G....R 12 FLOW 17B BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-17B-007 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the Emergency Condenser Makeup tanks.17B PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-17B-007 RA TURBINE FIRE PUMP run as needed to supply fire water to Emergency Condenser Makeup tanks.NMPII April 2013 Page G-34 I I NMP1, April 2013 Page G-34 I Constellation Energy Nuclear Group Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Component Component Description Actions VFDR RA/PCS Area 17B VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-11 is locally throttled to control VFDR-1 78-004 RA EMERGENCY CONDENSERS makeup to EC's 121 & 122 to provide a RCS heat sink for decay heat removal.17B VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-1 7B-003 RA EMERGENCY CONDENSERS makeup to EC's 111 & 112 to provide a RCS heat sink for decay heat removal.18 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-18-011 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the Emergency Condenser Makeup tanks.18 FCV-39-15 EMERGENCY CONDENSER FCV-39-15 operated locally via the VFDR-18-005 RA STEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory loss CONTROL VALVE -FLOW flow path from the EC's 111 & 112 CONTROL VALVE 11 STEAM RCS return path drain line.LINE DRAIN 18 FCV-39-16 EMERGENCY CONDENSER FCV-39-16 operated locally via the VFDR-18-004 RA STEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory loss CONTROL VALVE -FLOW flow path from the ECs 121 & 122 RCS CONTROL VALVE 12 STEAM return path drain line.LINE DRAIN 18 IV-01-04 24" AIR OPERATED (ANGLE) AIr vented manually at MSIV IV-01-04 VFDR-18-002 RA ISOLATION VALVE WITH to dose valve to prevent inventory loss.SOLENOID VALVES -MAIN STEAM OUT ISOLATION VALVE 4 18 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-18-011 RA TURBINE FIRE PUMP run as needed to supply fire water to Emergency Condenser Makeup tanks.18 VLV-05-31 LOOP 11 VENT LINE MANUAL VLV-05-31 is operated locally via the VFDR-18-003 RA ISOLATION VALVE DOWN hand wheel to isolate an inventory loss STREAM OF 05-01R AND 05-11 flow path from the ECs 111 & 112 RCS steam supply vent line.18 VLV-05-32 LOOP 12 VENT LINE MANUAL VLV-05-32 is operated locally via the VFDR-18-003 RA ISOLATION VALVE hand wheel to isolate an inventory loss DOWNSTREAM OF 05-04R AND flow path from the ECs 121 & 122 RCS 05-12 steam supply vent line.NMPI, April 2013 Page G-35 I!NMP1, April 2013 Page G-35 I Constellation Energy Nuclear Group Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RA/PCS 18 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-18-008 RA EMERGENCY CONDENSERS makeup to EC's 121 & 122 to provide a RCS heat sink for decay heat removal.18 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-18-008 RA EMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide a RCS heat sink for decay heat removal.4-9 aV-TO-43 11 AIR OpRN..ATE 91LOC.K,, 1V 79 53 is op-.ated locally to preo".do VFDR 49993 RA~~I~~j~~IV RBL ILTGI.I~ liAg Vtr to the 89G pumps and a TSH DOWNl COG'ING hMat sink for tho ROG HXc.49 FV 3. CID9 8" OPE.RATED FCV 33 09 is operated locally to VFDR40-0Q9 RA FLO 1COTRO VALVE eantrol flowi t-hrough SDC HX 38 132 to l SHUTDO0WN COOLINAG HEAT rogulato RCS- ca-l d-ewim Fate.ECHANG;R 441 FLOWA-44D FGV as 11 i "DAW r-MP..;..OP-PR.TED FC-3R IQ 19 BV-100-8 DIOPH111&2A-OE TD 2 CV-8 an i V-100-9are locally to VFDR-19-007 RA FLOW- C2NUTROL VALVES openedl flow henough SDc HX 3r 120 to GOH I-T 0DOW NCOOLI61N G HEA rogulate RCS see' down Fate.E-XCHANG-ER 12 FLOW EmrgncronenerMkep ans 9 I-108-02 MOETOR OPERIATECD IV 3P 02 is mpatnd lcally sta rte VFDR49-0076 RA SHU1-TDOWNA9A COOIN -OTLET- hand vpheel to octa-blich a- SDC suctio 201ID IfOATION' VALVE flev~ath from the RCS to acamplish________________________docay heat rmoeval.19 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-1 00-68 and BV-1 00-69 are locally VFDR-19-007 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the_________

______________________Emergency Condenser Makeup tanks.19 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-19-007 RA TURBINE FIRE PUMP run as needed to supply fire water to Emergency Condenser Makeup tanks.19 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-19-001 RA EMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide a RCS heat sink for decay heat removal.NMPI, April 2013 Page G-36 I m NMPI, April 2013 Page G-36 I Constellation Energy Nuclear Group Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RA/PCS 19 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-19-001 RA EMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide a RCS heat sink for decay heat removal.20 BV- 44-- " AIR 0OPE6RATED

' LOCKING BV. 70 53 62 epwrted !oly to pwid. VFDR RA V.ALkE R6LCINE VLVE- coo1ing W011F to9 the 69C pumpcD and a~O H NTAD l NCOO' ING heat ank for the 20C

  • 20 FGV 38-09 92 DLIA.PI.IRAGM OPE6RATE-D FCY 38 00 is oporotod locally to WFOR2-20-QG RA O CONITROL ,AI AVE cntral fow throuh C HX L 8 4126 tc 9HhDOW COOL ING HEAT rcgulato RC- coo' dwoni Fato 4 4 rI /60A CIZNITROL VAI 20._._G 8 4.. " I, A U. dA.P.R.. OPO ..TE' D r,-, uateC l do fl r. .2Q RA ErXCH.ANIG-E-R 412 FLOW9 20 IV 38 02 AMOTR OPERATED IV 38 02 is ep-r.t.d lea!.ly Y.1, "h. VFDR 20 00Q1 RA SHUTi'lDOWNti COOL 0-1ING OUTL6ET hand- v4,col to- e ctoblisch a 2 SDC-A Gcien OUTMID ....ATION VALVE fipath f., om t-. .RCS to ase.m p!lih__________

_____________decoy heat romovi2'.20 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-20-008 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the Emergency Condenser Makeup tanks.20 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-20-008 RA TURBINE FIRE PUMP run as needed to supply fire water to Emergency Condenser Makeup tanks.20 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-11 is locally throttled to control VFDR-20-002 RA EMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide a RCS heat sink for decay heat removal.20 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-20-002 RA EMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide a RCS heat sink for decay heat removal.NMPI, AprIl 2013 Page G-37 I I NMPI, April 2013 Page G-37 I Constellation Energy Nuclear Group Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Component Component Description Actions VFDR RA/PCS Area 2 RV4 53 14" AIR OPERATED BLOC81 "KING W 70 53 he oporat.d lceally to pr"ldo VFDR 24 094 RA I*)AjI~ BCC ILE VLVE coln -to the SDC pumps and a TO )HTOWN COOLING hat cA~k. forf the 89G HXs.24- FGV--8-0A 8" DIA-PHRPAG1.2..

OPERA-TED FCV 38 09 is oporated locally to VFW2.-4-006 RA FOW C-ONTROL VALVE c...ntrl Amow thruh--..*

6DC HX 38 125 te SHUTDOW COINGQ HEAT rgulato RCS cool do;A rato.E-=CH.AGR 11 FLOW 24- FGV s .4. 8 .DIAPHR.A.GM..

OPER-A.TE-D FOY 389 2. 4. oporoted Wcally R, 24 06 RA FL WCOTROL VALVE contro flowA. thrOu6gh SIX HX 38 4120 to SHUTDON COOLING I-EA rogulato RCS eeol devm Fate.EX=.CHANIGQrER 13 FL-OW 24- IV 38 02 OTROPERATED IV 38 02 Ic oporated locally ýAa toVFDR 21 004- RA 9HU TDO WNv C , I OUTLE.T hand .ea. l tno .ctabl!h a SDC ScUotro OU1 TSID IOR 'IN'.A flwopath from the RCS to accomplich

________________________decay heat reomwal, 21 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-21-008 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the Emergency Condenser Makeup tanks.21 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-21-008 RA TURBINE FIRE PUMP run as needed to supply fire water to Emergency Condenser Makeup tanks.21 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-21-002 RA EMERGENCY CONDENSERS makeup to EC's 121 & 122 to provide a RCS heat sink for decay heat removal.21 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-21-002 RA EMERGENCY CONDENSERS makeup to EC's 111 & 112 to provide a RCS heat sink for decay heat removal.22 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-22-008 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the Emergency Condenser Makeup tanks.22 FCV-39-16 EMERGENCY CONDENSER FCV-39-16 operated locally via the VFDR-22-004 RA STEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory loss CONTROL VALVE -FLOW flow path from the EC's 121 & 122 CONTROL VALVE 12 STEAM RCS return path drain line.lNMP1, April 2013 Page G-38 I Constellation Enerav Nuclear Group Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RA/PCS LINE DRAIN 22 IV-01-04 24- AIR OPERATED (ANGLE) Air vented manually at MSIV IV-01-04 VFDR-22-002 RA ISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAIN STEAM OUT ISOLATION VALVE 4 22 PMP-1 00-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-22-008 RA TURBINE FIRE PUMP run as needed to supply fire water to Emergency Condenser Makeup tanks.22 VLV-05-32 LOOP 12 VENT LINE MANUAL VLV-05-32 is operated locally via the VFDR-22-003 RA ISOLATION VALVE hand wheel to isolate an inventory loss DOWNSTREAM OF 05-04R AND flow path from the ECs 121 & 122 05-12 RCS steam supply vent line.22 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-11 is locally throttled to control VFDR-22-007 RA EMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide a RCS heat sink for decay heat removal.22 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-22-007 RA EMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide a RCS heat sink for decay heat removal.23 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-100-68 and BV-100-69 are locally VFDR-23-008 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the Emergency Condenser Makeup tanks.23 FCV-39-16 EMERGENCY CONDENSER FCV-39-16 operated locally via the VFDR-23-003 RA STEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory loss CONTROL VALVE -FLOW flow path from the ECs 121 & 122 RCS CONTROL VALVE 12 STEAM return path drain line.LINE DRAIN 23 IV-01-04 24- AIR OPERATED (ANGLE) AIr vented manually at MSIV IV-01-04 VFDR-23-002 RA ISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAIN STEAM OUT ISOLATION VALVE 4 NMPI, AprIl 2013 Page G-39 I I NMP1, April 2013 Page G-39 I Constellation Enerav Nuclear GrouD Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RA/PCS 23 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-23-008 RA TURBINE FIRE PUMP run as needed to supply fire water to Emergency Condenser Makeup tanks.23 VLV-05-32 LOOP 12 VENT LINE MANUAL VLV-05-32 is operated locally via the VFDR-23-004 RA ISOLATION VALVE hand wheel to isolate an inventory loss DOWNSTREAM OF 05-04R AND flow path from the ECs 121 & 122 RCS 05-12 steam supply vent line.23 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-23-006 RA EMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide a RCS heat sink for decay heat removal.23 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-23-006 RA EMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide a RCS heat sink for decay heat removal.24 BV-100-68 LOOP 111 & 112 AND LOOP 121 BV-1 00-68 and BV-1 00-69 are locally VFDR-24-007 RA BV-100-69

& 122 MANUAL VALVES opened to periodically refill the Emergency Condenser Makeup tanks.24 FCV-39-15 EMERGENCY CONDENSER FCV-39-15 operated locally via the VFDR-24-003 RA STEAM LINE DRAIN PRESSURE hand wheel to isolate an inventory loss CONTROL VALVE -FLOW flow path from the ECs 111 & 112 RCS CONTROL VALVE 11 STEAM return path drain line.LINE DRAIN 24 IV-01-04 24" AIR OPERATED ( ANGLE) Air vented manually at MSIV IV-01-04 VFDR-24-002 RA ISOLATION VALVE WITH to close valve to prevent inventory loss.SOLENOID VALVES -MAIN STEAM OUT ISOLATION VALVE 4 24 PMP-100-02 DIESEL DRIVEN VERTICAL PMP-100-02 is manually started and VFDR-24-007 RA TURBINE FIRE PUMP run as needed to supply fire water to Emergency Condenser Makeup tanks.24 VLV-60-11 MAKEUP VALVE TO LOOP 12 VLV-60-1 1 is locally throttled to control VFDR-24-004 RA EMERGENCY CONDENSERS makeup to ECs 121 & 122 to provide a RCS heat sink for decay heat removal.24 VLV-60-12 MAKEUP VALVE TO LOOP 11 VLV-60-12 is locally throttled to control VFDR-24-004 RA EMERGENCY CONDENSERS makeup to ECs 111 & 112 to provide a RCS heat sink for decay heat removal.NMPI, April 2013 Page G-40 I I NMPI, April 2013 Page G-40 I Constellation Energy Nuclear Group Attachment G -Recovery Actions Transition Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions VFDR RA/PCS 24 VLV-93-13 12" GATE VALVE -121 VLV-93-13 dosed locally via the hand VFDR-24-006 RA CONTAINMENT SPRAY RAW wheel to isolate CTSRW flow to WATER PUMP DISCHARGE Containment Spray Header #11.VALVE NMPArl21PaeG4

!NMPI, April 2013 Page G-41 I

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