IR 05000247/2004005

IR 05000247-04-005 on 03/01/2004 - 03/19/2004; Indian Point Nuclear Generating Unit No. 2; Triennial Fire Protection Inspection. In-office Inspection follow-up Activities Occurred Through April 16, 2004

ML042330248
Person / Time
Site:	Indian Point
Issue date:	08/20/2004
From:	Rogge J NRC/RGN-I/DRS/EB
To:	Dacimo F Entergy Nuclear Operations
References
IR-04-005
Download: ML042330248 (28)

Text

ust 20, 2004

SUBJECT:

INDIAN POINT NUCLEAR GENERATING UNIT NO. 2 - NRC TRIENNIAL FIRE PROTECTION INSPECTION REPORT 05000247/2004005

Dear Mr. Dacimo:

On March 19, 2004, the US Nuclear Regulatory Commission (NRC) completed a triennial fire protection inspection at your Indian Point Unit 2 facility. The enclosed report documents the inspection findings, which were discussed on July 20, 2004, with you and other members of your staff.

This inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations, and with the conditions of your license. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

Based on the results of this inspection no findings of significance were identified. However, licensee-identified violations which were determined to be of very low safety significance are listed in Section 40A7 of this report. If you contest these non-cited violations, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region 1; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at Indian Point Unit 2.

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARs) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Mr. Fred Dacimo 2 We appreciate your cooperation. Please contact me at (610) 337-5146 if you have any questions regarding this letter.

Sincerely,

/RA/

John F. Rogge, Chief Electrical Branch Division of Reactor Safety Docket Nos. 50-247 License Nos. DPR-26

Enclosure:

NRC Inspection Report 05000247/2004005 w/Attachment: Supplemental Information

REGION 1 Docket No: 50-247 License No: DPR-26 Report No: 05000247/2004005 Licensee: Entergy Nuclear Northeast Facility: Indian Point Nuclear Generating Unit No. 2 Location: Buchanan, NY Inspection Dates: March 1 - 19, 2004 (In office inspection follow-up through April 16, 2004)

Inspectors: Keith Young, Senior Reactor Inspector, DRS (Team Leader)

Larry Scholl, Senior Reactor Inspector, DRS Timothy OHara, Reactor Inspector, DRS Jonathan Lilliendahl, Reactor Inspector, DRS (Observer)

Approved by: John F. Rogge, Chief Electrical Branch Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS IR 05000247/2004005; 03/01/2004 - 03/19/2004; Indian Point Nuclear Generating Unit No. 2; Triennial Fire Protection Inspection. In-office inspection follow-up activities occurred through April 16, 2004.

The report covered a two week team inspection by specialist inspectors. The team identified one unresolved item. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.

A. NRC-Identified Findings No findings of significance were identified.

B. Licensee-Identified Violations Violations of very low safety significance, which were identified by Entergy Nuclear Northeast have been reviewed by the team. Corrective actions taken or planned by Entergy Nuclear Northeast have been entered into their corrective action program.

These violations and corrective actions are listed in Sections 4OA3 and 4OA7 of this report.

ii Enclosure

Report Details Background This report presents the results of a triennial fire protection inspection conducted in accordance with NRC Inspection Procedure (IP) 71111.05, Fire Protection. The objective of the inspection was to assess whether Entergy Nuclear Northeast has implemented an adequate fire protection program and that post-fire safe shutdown capabilities have been established and are being properly maintained at the Indian Point Nuclear Generating Unit No. 2. The following fire zones (FZ) and a fire area (FA) were selected for detailed review based on risk insights from the Indian Point Nuclear Generating Station, Unit 2, Individual Plant Examination (IPE) and Individual Plant Examination of External Events (IPEEE):

C Fire Zone 10 - Diesel Generator Building C Fire Zone 14 - 480 V Switchgear Room C Fire Zone 15 - Central Control Room C Fire Zone 32A - Electrical Tunnel C Fire Area J - Turbine Building This inspection was a reduced scope inspection in accordance with the March 6, 2003, revision to IP 71111.05, Fire Protection. Issues regarding equipment malfunction due to fire-induced failures of associated circuits were not inspected. Criteria for review of fire-induced circuit failures are currently the subject of a voluntary industry initiative. The definition of associated circuits of concern used was that contained in the March 22, 1982, memorandum from Mattson to Eisenhut which clarified the requests for information made in NRC Generic Letter (GL) 81-12.

1. REACTOR SAFETY Cornerstones: Initiating Events, Mitigating Systems 1R05 Fire Protection 1. Programmatic Controls a. Inspection Scope During tours of the facility, the team observed the material condition of fire protection systems and equipment, the storage of permanent and transient combustible materials, and control of ignition sources. The team also reviewed the procedures that controlled hot-work activities and combustibles at the site. These reviews were accomplished to ensure that the Entergy Nuclear Northeast was maintaining the fire protection systems, controlling hot-work activities, and controlling combustible materials in accordance with the Updated Final Safety Analysis Report (UFSAR)/Fire Protection Program Plan (FPPP), administrative procedures and other fire protection program procedures.

Enclosure

b. Findings No findings of significance were identified.

2. Fire Area Boundaries and Barriers a. Inspection Scope The team walked down accessible portions of the selected fire areas to observe material condition and the adequacy of design of fire area boundaries, fire doors, and fire dampers. The team reviewed engineering evaluations, as well as surveillance and functional test procedures for selected items. The team also reviewed the licensee submittals and NRC safety evaluation reports (SERs) associated with fire protection features at Indian Point Unit 2. Additionally, the team reviewed the design and qualification testing of selected barriers and reviewed surveillance procedures for structural fire barriers, penetration seals and structural steel. These reviews were performed to ensure that the passive fire barriers were properly maintained and met the licensing and design bases as described in the licensee submittals, NRC SERs, the Indian Point Unit 2, UFSAR and the FPPP.

b. Findings No findings of significance were identified.

3. Post-Fire Safe Shutdown Lighting and Communications a. Inspection Scope The team observed the placement and aim of eight-hour emergency lights throughout the selected fire areas to evaluate their adequacy for illuminating access and egress pathways and any equipment requiring local operation for post-fire safe shutdown. The team also reviewed preventive maintenance procedures and various documents, including the vendor manual and surveillance tests, to determine if adequate surveillance testing and periodic battery replacements were in place to ensure reliable operation of the emergency lights.

The team reviewed radio repeater location and power sources to ensure fire brigade and operator communications could be maintained for fire fighting and post-fire safe shutdown conditions.

b. Findings No findings of significance were identified.

Enclosure

4. Fire Detection Systems and Equipment a. Inspection Scope The team reviewed the adequacy of the fire detection systems in the selected plant fire areas. This included a walkdown of the systems and review of the type of installed detectors as shown per location drawings. The team also reviewed licensee submittals, engineering evaluations and the NRC SERs associated with the selected fire areas.

These reviews were performed to ensure that the fire detection systems for the selected fire areas were installed in accordance with the design and licensing bases of the plant.

Additionally, the team reviewed fire detection surveillance procedures to determine the adequacy of the fire detection component testing and to ensure that the detection system would function as required.

b. Findings No findings of significance were identified.

5. Fixed Fire Suppression Systems a. Inspection Scope Sprinkler Systems and Equipment The team reviewed the adequacy of the pre-action sprinkler system for FZ 32A, Electrical Tunnel, and the wet pipe sprinkler system in FZ 10, Diesel Generator Room, by performing a walkdown and observing the material condition of the system.

Additionally, the team reviewed the design and installation specifications, NFPA 13, Standard for the Installation of Sprinkler Systems, installation drawings, the adequacy of surveillance procedures and hydraulic calculations. These reviews were performed to ensure that the sprinkler system met the design and licensing bases as described in the licensee submittals, NRC SERs and the UFSAR/FPPP, and that the systems could perform their intended function in the event of a fire in their respective areas.

b. Findings No findings of significance were identified.

6. Manual Fire Suppression Capability a. Inspection Scope The team walked down selected standpipe systems and observed portable extinguishers to determine the material condition of the manual fire fighting equipment and verify locations as specified in the pre-fire plans and fire protection program documents. The team reviewed electric and diesel fire pump flow and pressure tests to Enclosure

ensure that the pumps were meeting their design requirements. The team also reviewed the fire main loop flow tests to ensure that the flow distribution paths were able to meet the design requirements. The team inspected the fire brigades protective ensembles, self-contained breathing apparatus (SCBA), and various fire brigade equipment to determine operational readiness for fire fighting.

The team reviewed pre-fire plans and smoke removal strategies for the selected fire areas to determine if appropriate information was provided to fire brigade members and plant operators to identify safe shutdown equipment and instrumentation, and to facilitate suppression of a fire that could impact safe shutdown.

The team performed in-plant walkdowns to evaluate the physical configuration of electrical raceway and safe shutdown components in the selected fire areas to determine whether water from an inadvertent fire suppression system pipe rupture or from manual fire suppression activities in the selected areas could cause damage that could inhibit the ability to safely shut down the plant.

The team reviewed fire brigade initial training and continuing training course materials to verify appropriate training was being conducted for the station firefighting personnel.

Additionally, the team reviewed selected fire drills and critiques to ensure that drills were being conducted in risk significant areas.

The team reviewed the qualifications of several fire brigade leaders and members to ensure that they had met and maintained the requirements to be qualified fire brigade members.

b. Findings No findings of significance were identified.

7. Post-Fire Safe Shutdown Capability a. Inspection Scope The team reviewed the fire response procedures, alarm response procedures, operating procedures for the selected fire areas, and the FPPP to evaluate the methods and equipment used to achieve hot shutdown following a fire. The team also reviewed piping and instrumentation drawings (P&ID) for post-fire safe shutdown systems to identify required components for establishing flow paths, to identify equipment required to isolate flow diversion paths, and to verify appropriate components were properly evaluated and included in the safe shutdown equipment list. The team also reviewed selected alternate safe shutdown components and their control circuits to ensure that proper isolation was provided for alternate safe shutdown capability, in the event of a fire affecting the control room, cable tunnel or the 480 V switchgear room. The team performed field walkdowns to evaluate the protection of the equipment from the effects of fires.

Enclosure

A post-fire safe shutdown procedure for the selected areas was also reviewed to determine if appropriate information was provided to plant operators to identify protected equipment and instrumentation and if recovery actions specified in the post-fire safe shutdown procedure considered manpower needs for performing required actions. The team also reviewed training lesson plans for the alternative shutdown procedures, discussed training with licensed operators, reviewed selected alternate safe shutdown equipment tests and calibrations, reviewed the adequacy of shift manning, and evaluated the accessibility of the alternative shutdown operating stations and required manual action locations.

The specific alternate safe shutdown procedure reviewed was 2-AOP-SSD-1, Control Room Inaccessibility Safe Shutdown Control, Revision 1. A procedure walkdown was performed for this procedure. The walkdown was performed by an operations crew and focused primarily on the portion of the procedure associated with achieving stable hot shutdown conditions. Plant operators were accompanied by NRC team members during the walkdown and the approximate time for critical steps, such as establishing makeup flow to the reactor vessel, aligning the alternate safe shutdown system (ASSS)

to supply electrical power to safe shutdown loads and establishing component cooling water (CCW) for reactor coolant pump (RCP) seal cooling were noted and evaluated to assess the ability of the operators to maintain plant parameters within procedural limits.

b. Findings Introduction. The team identified and unresolved item concerning the ability of the alternate safe shutdown procedure 2-AOP-SSD-1, Control Room Inaccessibility Safe Shutdown Control, Revision 1, to ensure a safe shutdown of the plant for a severe fire in the 480 volt switchgear room. Specifically, the procedure may not have precluded an extended loss of reactor coolant pump seal injection flow which could have led to RCP leakage above the Westinghouse Owners Group (WOG) guidance of 21 gallons per minute after 13 minutes. This issue will remain unresolved pending further NRC review of information regarding RCP seal performance following a loss of all seal cooling event.

Description. Following review and a walkdown of Revision 1 of procedure AOP-SSD-1, the team noted the following issues that reduced its efficiency. The teams observations were as follows:

a. The procedure did not prioritize actions to ensure prompt restoration of electrical power and RCP seal cooling via seal injection and/or CCW flow to the seal coolers.

Enclosure

b. After GT-1 is started, the procedure verifies if CCW and/or a charging pump is operating. If they are not running, instead of starting the pumps on ASSS power (GT-1), the procedure had operators preparing to start a safety injection (SI)

pump. The SI pump requires casualty cable installation and would not be immediately available.

c. The step to start a charging pump on ASSS power came after initiating the SI pump actions.

d. Based on their training, the operators assumed RCP seal damage may occur after 15 minutes without RCP seal injection and, therefore, the procedure focus on SI pump recovery seemed to make sense to them.

e. On March 19, 2004, Entergy Nuclear Northeast discovered that GT-1 would not have started locally because of a failed op-trend computer in the control room.

The team expressed these and other concerns to plant personnel. Entergy Nuclear Northeast initiated CR-2004-1445 to complete a recovery plan for the identified procedural concerns. The recovery plan included implementing compensatory measures, revising the FPPP, revising procedure AOP-SSD-1 and performing a calculation to ensure the ASSS procedure steps could be performed to meet the performance goals for 10 CFR 50, Appendix R. Additionally, Entergy Northeast performed timeline validations of the revised procedure.

Based on the observed walkdown of Revision 1 of procedure AOP-SSD-1, the team estimated that more than 60 minutes could elapse without charging or CCW before the ASSS would be aligned to provide electrical power to these systems. Current WOG guidance suggest that this prolonged loss of electrical power, that supplies RCP seal cooling, could cause increased seal leakage after 13 minutes. After some time, RCP seal leakage could become excessive. If RCP seal cooling is not established within 13 minutes following the loss of electrical power, current WOG guidance suggests that it should not be established prior to performing a cooldown to protect the seals.

Additionally, the FPPP references NUREG/CR-2934, Review and Evaluation of the Indian Point Probabilistic Safety Study, which states that seal failure is assumed to occur following failure of the redundant means of providing cooling (i.e., charging and the component cooling system) and is predicted to lead to significant seal leakage at 30 minutes. The Safety Evaluation of Topical Report WCAP-15603, Revision 1, WOG 2000 Reactor Coolant Pump Seal Leakage Model For Westinghouse PWRs, indicates that leakage of RCP seal could increase to more than 21 gpm after 13 minutes.

The identified issue could adversely impacted the efficiency of Entergy Nuclear Northeasts safe shutdown strategy and could have prevented achieving and maintaining hot shutdown in the post-fire environment. Because this issue could affected the reactor safety mitigating system cornerstone objective, the issue could be greater than minor. The safety significance of this finding was not evaluated due to further NRC review of RCP seal leakage issues.

Enclosure

It appeared that Entergy Nuclear Northeast did not provide an alternate safe shutdown procedure AOP-SSD-1, Revision 1, that would efficiently protect RCP seals from potentially leaking more than the 21 gpm after 13 minutes per NRC safety evaluation guidance and the guidance provided in NUREG/CR-2934, that excessive RCP seal leakage could occur after 30 minutes. Revision 4 of the alternate safe shutdown procedure appeared to incorporate several improvements that increased the efficiency of accomplishing the necessary safe shutdown tasks. The team concluded that the identified issue concerning potential effects on alternate safe shutdown for RCP seals is an unresolved item (URI) pending further NRC review of information regarding RCP seal performance following a loss of all seal cooling event. This issue is identified as URI 05000247/2004005-01, RCP Seal Cooling for Alternate Safe Shutdown Procedure Strategies.

8. Safe Shutdown Circuits a. Inspection Scope The team reviewed power and control cable routing for a sample of components required for post-fire safe shutdown to determine if the cables were properly evaluated as part of the safe shutdown analyses in the UFSAR/FPPP. The team performed a walkdown of accessible selected alternate safe shutdown system cables to ensure cables were not routed in the inspected fire areas. Additionally, the team reviewed cable routing for selected safe shutdown components. This was accomplished by reviewing cable routing drawings and performing infield verification of accessible cables to ensure they were not routed in the inspected fire areas.

The team reviewed electrical fuse and circuit breaker coordination studies to ensure equipment needed to conduct post-fire safe shutdown activities would not be impacted due to a lack of coordination. The team confirmed that coordination studies had addressed multiple faults due to fire. The team also reviewed the electrical isolation capability of selected equipment needed for post-fire safe shutdown to ensure that such equipment could be operated locally if needed.

Additionally, the team reviewed calculations and procedures required to ensure minimum operation voltages were being supplied to alternate safe shutdown equipment.

Due to the issuance of Change Notice 00-020 to Inspection Procedure 71111.05, Fire Protection, the team did not review associated circuit issues during this inspection.

This change notice has suspended this review pending completion of an industry initiative in this area.

Enclosure

b. Findings No findings of significance were identified.

4. OTHER ACTIVITIES 4OA2 Identification and Resolution of Problems 1. Corrective Actions for Fire Protection Deficiencies a. Inspection Scope The team reviewed the open corrective maintenance work orders for fire protection and safe shutdown equipment, selected condition reports (CRs) for fire protection and safe shutdown issues and the fourth quarter (2002) Unit 2 fire protection systems health report to evaluate the prioritization for resolving fire protection related deficiencies and the effectiveness of corrective actions. The team also reviewed recent Quality Assurance (QA) audits and self-assessments of the fire protection program to determine if the licensee was identifying program deficiencies and implementing appropriate corrective actions.

b. Findings No findings of significance were identified.

4OA3 Event Follow-up 1. (Closed) LER 05000247/2003001-00, Plant is an Unanalyzed Condition due to Cable Routing Non-Compliance with Appendix R Separation Criteria On February 6, 2003, Entergy Nuclear Northeast identified that the routing of charging pump control cables did not meet the cable separation criteria specified in 10 CFR 50, Appendix R, section III.G. On February 12, 2003, Entergy Nuclear Northeast further identified that the routing of the power supply cables for the six service water pumps did not meet the separation criteria of 10 CFR 50, Appendix R, section III.G. These issues were identified during validation and re-baselining of the10 CFR 50, Appendix R, post-fire safe shutdown analysis. It was determined that the identified cables would be vulnerable to fire damage in the fire areas that the cables are routed. Charging pump cables would be vulnerable to a fire in the primary auxiliary building (fire area F) and service water pump cables would be vulnerable to a fire in the turbine building (fire area J). Entergy Nuclear Northeast determined the apparent cause of these issues was lack of rigor applied to the engineering analysis that developed the post fire safe shutdown analysis. Specifically, the engineering analysis lacked sufficient detail and/or support documentation to justify the original design configuration. Corrective actions included implementation of compensatory actions upon discovery, completion of re-baselining of the Appendix R analysis and implementation of plant modifications (planned Enclosure

modifications) to resolve non-compliance issues with Appendix R. These findings are more than minor because they had a credible impact on safety, in that if a fire had occurred in the identified areas, systems used to shut down the plant could have been challenged. These findings affected the Mitigating Systems Cornerstone and were considered to have a very low safety significance (Green). The Region I, Senior Risk Analyst reviewed and agreed with Entergy Nuclear Northeasts determination of risk associated with the charging pump issue. This risk was determined to be of very low significance. The team used Appendix F of the fire protection significance determination process (SDP) to determine the risk associated with the service water issue and the risk was determined to be of very low safety significance. In both cases, the ASSS system was available to shutdown the plant in the event that a fire challenged the systems.

Additionally, developing a credible fire in the identified areas to damage all cables is of low probability. These licensee identified findings involved violations of 10 CFR 50, Appendix R, section III. G., cable separation criteria. This LER is closed.

4OA5 Other (Closed) URI 05000247/2003003-01: Lack of Cable Separation in Fire Areas F and J, Postulated Fire Compromising Associated Circuits This unresolved item (URI) was opened in NRC Inspection Report 05000247/200303.

The team reviewed this URI and the associated LER 05000247/2003001-00. This URI is closed based on the disposition of LER 05000247/2003001-00 which is described in section 4OA3 of this report.

4OA6 Meetings, including Exit The team presented their preliminary inspection results to Mr. C. Schwarz, GM Plant Operations, and other members of the Indian Point Nuclear Generating Unit No. 2 staff at a debrief meeting on March 18, 2004, and to Mr. F. Dacimo, Site Vice President, at the exit meeting on July 20, 2004. The findings were acknowledged by Entergy Nuclear Northeast. The inspectors confirmed that proprietary information was not provided or examined during the inspection.

4OA7 Licensee-Identified Violations The following violations of very low safety significance (Green), which were identified by the licensee, were reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the Entergy Nuclear Northeasts corrective action program.

Enclosure

• License Condition 2K requires Entergy Nuclear Northeast to implement and maintain all provisions of the NRC approved fire protection program as described in the UFSAR for the facility and as approved in the SERs dated November 30, 1977, February 3, 1978, January 31, 1979, October 31, 1980, August 22, 1983, March 30, 1984, October 16, 1984, September 16, 1985, November 13, 1985, March 4, 1987, January 12, 1989, and March 26, 1996. By SER dated March 30, 1984, the NRC completed its review of the licensees submittal (and exemptions) and concluded that the licensee was in compliance with 10 CFR 50.48, Appendix R, Sections III.G, and Section III.L. 10 CFR 50, Appendix R Section III. G. 1.(a) & (b) requires fire protection features be provided for systems, structures, and components important to safe shutdown. These features shall be capable of limiting fire damage so that: (a) One train of systems necessary to achieve and maintain hot shutdown conditions from either the control room or emergency control station(s) is free from fire damage; and (b)

Systems necessary to achieve and maintain cold shutdown from either the control room or emergency control station(s) can be repaired within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Contrary to this, the Entergy Nuclear Northeast identified, via CR-IP2-2003-06834 and CR-IP2-2003-02428, that the pneumatic tubing on the steam generator atmospheric steam dump valves contained soldered copper fittings which could be vulnerable to failure if exposed to the high temperature environment of a postulated fire. Such a failure could result in a common-mode failure of all pneumatic lines supplying all four steam generator atmospheric steam dump valves. This failure could disable the atmospheric steam dump valve positioners from both instrument air and nitrogen, thus disabling all steam generator dump valves which are credited for use during alternate safe shutdown system operation for post fire safe-shutdown events. This licensee identified finding is of very low safety significance because the likelihood of occurrence of a fire that could damage all tubing in this area is small, there were no significant combustibles in the area and no loss of post-fire safe-shutdown capability occurred. Additionally, compensatory measures were established pending resolution of this issue. This issue screened to green in phase I of the reactor SDP, MC0609, Appendix A, because it did not involve the total loss of any safety function that would have contributed to identified external event initiated core damage accident sequences.

• License Condition 2K requires Entergy Nuclear Northeast to implement and maintain all provisions of the NRC approved fire protection program as described in the UFSAR for the facility and as approved in the SERs dated November 30, 1977, February 3, 1978, January 31, 1979, October 31, 1980, August 22, 1983, March 30, 1984, October 16, 1984, September 16, 1985, November 13, 1985, March 4, 1987, January 12, 1989, and March 26, 1996. By SER dated March 30, 1984, the NRC completed its review of the licensees submittal (and exemptions) and concluded that the licensee was in compliance with 10 CFR 50.48, Appendix R, Sections III.G, and Section III.L. 10 CFR 50, Appendix R, Section III. G. 2., requires, except as provided for in paragraph G.3 of this section, where cables or equipment, including associated non-safety circuits that Enclosure

could prevent operation or cause maloperation due to hot shorts, open circuits, or shorts to ground, of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment, one of the following means of ensuring that one of the redundant trains is free of fire damage shall be provided. Contrary to this, Entergy Nuclear Northeast identified, via CR-IP2-2003-01508, that control cables for the containment nitrogen header vent valve HCV-943 were vulnerable to fire-induced faults for fires in Fire Areas A and H. The potential fire-induced faults could cause spurious opening of the valve HCV-943, causing the nitrogen supply to all pneumatically-operated ASSS instruments to be lost. This failure could cause steam generator level, pressurizer level and pressurizer pressure indications to be lost which could disable the ASSS indications. This licensee identified finding is very low safety significance because the likelihood of occurrence of a fire that could damage control cables for these valves is small, procedural steps in the alternate safe shutdown procedure exist to re-enable the ASSS instruments and no loss of safety function occurred. Additionally, compensatory measures were established pending resolution of this issue. This issue screened to green in phase I of the reactor SDP, MC0609, Appendix A, because it did not involve the total loss of any safety function that would have contributed to identified external event initiated core damage accident sequences.

• License Condition 2K requires Entergy Nuclear Northeast to implement and maintain all provisions of the NRC approved fire protection program as described in the UFSAR for the facility and as approved in the SERs dated November 30, 1977, February 3, 1978, January 31, 1979, October 31, 1980, August 22, 1983, March 30, 1984, October 16, 1984, September 16, 1985, November 13, 1985, March 4, 1987, January 12, 1989, and March 26, 1996. By SER dated March 30, 1984, the NRC completed its review of the licensees submittal (and exemptions) and concluded that the licensee was in compliance with 10 CFR 50.48, Appendix R, Sections III.G, and Section III.L. 10 CFR 50, Appendix R, Section III. G. 2., requires, except as provided for in paragraph G.3 of this section, where cables or equipment, including associated non-safety circuits that could prevent operation or cause maloperation due to hot shorts, open circuits, or shorts to ground, of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment, one of the following means of ensuring that one of the redundant trains is free of fire damage shall be provided. Contrary to this, the Entergy Nuclear Northeast identified, via CR-IP2-2003-06341, that the control cables for charging pump suction path valves LCV-112B and LCV-112C were susceptible to spurious operation (closure). Closure of these valves could result in isolation of both normal and emergency suction paths to the charging pumps.

A fire in certain areas of the plant could result in damage to any running charging pumps at the start of the event. Potential pump damage could render the affected charging pumps inoperable from either normal or ASSS power sources, thus potentially disabling all charging capability which is credited in support of Enclosure

Appendix R performance criteria. This licensee identified finding is of very low safety significance because the likelihood of occurrence of a fire that could damage control cables for these valves is small and no actual loss of safety function occurred. Additionally, compensatory measures were established pending resolution of this issue. This issue screened to green in phase I of the reactor SDP, MC0609, Appendix A, because it did not involve the total loss of any safety function that would have contributed to identified external event initiated core damage accident sequences.

ATTACHMENT: SUPPLEMENTAL INFORMATION Enclosure

A-1 ATTACHMENT SUPPLEMENTAL INFORMATION KEY POINTS OF CONTACT Licensee Personnel Site Vice President C. Schwarz Admin.-GM Plant Operations V. Andreozzi Electrical Design Manager F. Bloise DBI Project P. Conroy Licensing Manager J. Comiotes Director Nuclear Safety Assurance G. Dahl Licensing K. Elliot IP2 Fire Protection Engineer M. Garofelo QA Supervisor P. Gropp DBI Project Manager T. Jones Licensing D. Leach Director Engineering L. Lubrano Electrical Maintenance R. Milici Design Engineering T. Orlando PCE Manager S. Petrosi Design System Manager J. Raffaele Electrical Design Engineering Supervisor J. Reynolds CA&A H. Robinson Electrical Design Engineer S. VanBuren Fire Chief J. Ventosa Site Operations Manager T. Williams IP2 Ops Manager S. Wilkie Fire Protection Engineer NRC Personnel W. Lanning, Director, Division of Reactor Safety J. Rogge, Chief, Electrical Engineering Branch P. Drysdale, Senior Resident Inspector, Indian Point Unit 2 M. Cox, Resident Inspector, Indian Point Unit 2 Attachment

A-2 LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED Opened 05000247/2004005-01 URI RCP Seal Cooling for Alternate Safe Shutdown Procedure Strategies.

Opened and Closed NONE Closed 05000247/2003003-01 URI Lack of Cable Separation in Fire Areas F and J, Postulated Fire Compromising Associated Circuits05000247/2003001-00, LER Plant in an Unanalyzed Condition due to Cable Routing Non-Compliance with Appendix R Separation Criteria Discussed NONE LIST OF DOCUMENTS REVIEWED Fire Protection Documents and Safety Evaluation Reports Indian Point Unit No. 2, Nuclear Generating Station, Individual Plant Examination of External Events, 1995 Indian Point Unit No. 2, Nuclear Generating Station, Individual Plant Examination Indian Point Unit 2, Fire Protection Program Plan, Rev. 9 & Rev. 10 Indian Point Unit 2, Nuclear Generating Station, Updated Final Safety Analysis Report, Rev. 14 Indian Point Unit No. 2, Fire Protection Safety Evaluation Report, January 31, 1979 Indian Point Unit No. 2, Safety Evaluation Report, Appendix R to 10CFR Part 50, Items III.G. and III.L., August 22, 1983 Indian Point Unit No. 2, Supplemental Safety Evaluation (SSE) for Appendix R to 10CFR50, Sections III.G. and III.L., March 30, 1984 Indian Point Unit No. 2, Safety Evaluation Report, Control Room Penetration Seals, September 16, 1985 Indian Point Unit No. 2, Safety Evaluation of Fire Protection Clarifications, January 12 1989 Indian Point Unit No. 2 Safety Evaluation Report Susceptibility of Safety-Related Systems To Flooding From Failure of Non-Category I Systems for IP2 Attachment

A-3 Indian Point Unit No. 2, Technical Exemption From The Requirements of 10CFR50 Appendix R For The Indian Point Nuclear Generating Unit No. 2, November 13, 1985 Indian Point Unit No. 2, System Description Number 32, Alternate Safe Shutdown System, Rev. 1 Indian Point Unit No. 2, Indian Point Unit 2 Pre-Fire Plans Design Basis Documents CYW DBD, Entergy Nuclear Northeast Indian Point 2 Design Basis Document For The Unit 2 City Water System, Rev. 0 IP2-DBD-221, Entergy Nuclear Northeast Indian Point 2 Nuclear Power Plant Water Supply, Distribution, Pumping Facilities, Water-Based Fire Suppression Systems and Hose Stations, Tab I, Rev. 0 IP2-DBD-221, Entergy Nuclear Northeast Indian Point 2 Nuclear Power Plant Fire Detection and Alarm Systems, Tab III, Rev. 0 IP2-DBD-221, Entergy Nuclear Northeast IP2 Nuclear Power Plant Appendix R Emergency Lighting System, TAB V, Rev. 0 IP2-DBD-222, Entergy Nuclear Northeast Indian Point Unit 2, Design Basis Document For Electrical Separation, Rev. 1 Procedures ENN-DC-161, Transient Combustible Program, Rev. 0 ENN-LI-102, Corrective Action Process, Rev. 0 ENN-LI-104, Self-Assessment and Benchmark Process, Rev. 4 IP-SMM-DC-901, IPEC Fire Protection Program Plan, Rev. 0 PC-R37, Alternate Safe Shutdown Instrument Calibration Check, Rev. 9 SAO-702, Control of Ignition Sources, Rev. 9 SAO-703, Fire Protection Impairment Criteria and Surveillance, Rev. 12 SAO-704, Removal/Reinstallation Fire Rated Assemblies, Rev. 5 SAO-705, Fire Watch/Fire Watch Tour, Rev. 10 SAO-711, Quality Assurance Requirements For Fire Protection Systems, Rev. 0 TNG-AD-16, Fire Protection Training Program, Rev. 6 TPD 407-QA, Fire Brigade Training, Rev. 1 2AOP-SSD-1, Control Room Inaccessibility Safe Shutdown Control, Rev. 1 2AOP-SSD-1, Control Room Inaccessibility Safe Shutdown Control, Rev. 2 2AOP-SSD-1, Control Room Inaccessibility Safe Shutdown Control, Rev. 4 2-OSAL 15.13, Abnormal Breaker Operation, Rev. 1 IP2 Maintenance Procedure Number BRK-B-004-N, ITE Type KC, Model G-Air Circuit Breaker - Non-Class "A", Rev. 9 IP2 Maintenance Procedure Number BRK-B-004-N, ITE Type KC, Model G-Air Circuit Breaker - Non-Class "A", Rev. 10 IP2 Equipment Reliability, Operational Instrument Calibration Procedure Number: ER-02, Rev. 0 (PI-1356)

Attachment

A-4 IP2 Maintenance Procedure, ITE Type KC, Model 6-AIR CIRCUIT BREAKER - NON-CLASS

"A", Procedure Number BKR-B-004-N, Rev. 9 IP2 Maintenance Procedure, ITE Type KC, Model 6-AIR CIRCUIT BREAKER - NON-CLASS

"A", Procedure Number BKR-B-004-N, Rev. 10 Surveillance Procedures LARP 10, Diesel Fire Pump System, Rev. 9 PC-R37, Alternate Safe Shutdown Instruments Surveillance Test, Rev. 9, completed October 20, 200010/20/00 PI-M4, Fire Brigade Equipment, Rev. 9, completed January 1, 2004 and January 29, 2004 PI-M5, Alternate Safe Shutdown System Cabinet Inventory, Rev. 14, completed January 4, 2004 and January 31, 2004 PI-M7, Fire Extinguisher Inspections, Rev. 25 PI-Q1, Fire Separation Barriers, Rev. 7 PT-A16, Electrical Tunnel Water Sprinkler System, Rev. 12 PT-A17A, Fire Hose Stations - Safe Shutdown, Rev. 6 PT-A17B, Fire Hose Stations, Rev. 6 PT-A22, Fire Loop Flow, Rev. 3, completed January 22, 1998 PT-A28, Emergency Light Battery Integrity, Rev. 8 PT-EM15, Diesel Generator Building Water Sprinkler System, Rev. 4 PT-EM28, Fire Dampers Operability, Rev. 5 PT-M34A, 11 Fire Main Booster Pump, Rev. 1, completed January 28, 2004 and February 25, 2004 PT-M34B, 12 Fire Main Booster Pump, Rev. 1 Completed January 29, 2004 and February 26, 2004 PT-M40, Diesel Fire Pump, Rev. 22 PT-M49A, Appendix R Emergency Lighting (Conventional), Rev. 16 PT-M49B, Appendix R Emergency Lighting (Nuclear), Rev. 10 PT-Q17A, Alternate Safe Shutdown Supply Verification to 21 AFP, Rev. 10, Completed August 11, 2003 PT-Q17B, Alternate Safe Shutdown Supply Verification to 23 CHP, Rev. 9, Completed August 7, 2003 & January 22, 2004 PT-Q17C, Alternate Safe Shutdown Supply Verification to 23 CHP, Rev. 9, Completed November 17,2003 PT-SA11, Diesel Generator Building Fire Detection System, Rev. 5 PT-SA12A, Ionization Type Smoke Detector (Conventional), Rev. 8 PT-SA12B, Ionization Type Smoke Detector (PAB), Rev. 6 PT-3Y15, Fire Loop Flow, Rev. 0, completed September 18, 2001 TST-PI-M5, Alternate Safe Shutdown System Cabinet Inventory, Rev. 14, completed March 27, 2003 TST-PT-EM19,Cable Spreading Room Halon System, Rev. 7, completed July 8, 2003 TST-PT-Q17A, Verify ASSS Supply To 21 AFP, Rev. 10, Completed October 19, 2003 &

November 13, 2003 Attachment

A-5 TST-PT-Q17B, Verify ASSS Supply To 23 CHP, Completed May 14, 2003 & October 23, 2003 W.O. NP-00-14142, Check N5143 In Accordance With I&C Work Step List, Completed July 5, 2000 W.O. IP2-01-21490, Perform ICPM-1346 ACS/CCW Return Charging Pump Fluid Drives FI-637, Completed July 18, 2003 W.O. IP2-01-21579, Perform ICPM-1293; 21, 22, 23, 24 RCP Seal Water Supply Flow, Completed May 22, 2002 W.O. IP2-03-14171, Child of IP2-02-33445 For PWT To Perform PT-Q17D for 12RW3-1M, Completed March 18, 2003 Fire Brigade Training Documents Fire Brigade Training Manual (including lessons)

IPEC Tour of Offsite Fire Departments, February 8, 2003 Offsite Fire Department Training, June 17, 2002 and August 18, 2003 Initial Fire Brigade Training Exams, 2003 Annual Fire Brigade Retraining Exams, 2003 2003/2004 Fire Brigade Record, February 29, 2004 Drills & Critiques No. 21 Battery Room, Completed February 15, 2004 and February 22, 2004 53 EL. Turbine Building - Bearing #2, Completed February 9, 2004 15 EL. Turbine Building - Seal Oil Vacuum Pump Unit 2, Completed September 23, 2003 and November 30, 2003 21 Aux. Boiler Feed Pump, Completed May 25, 2003 Insulation Storage Area, 33 EL. Unit 2, Completed January 19, 2003 QA Audits, Self-Assessments & System Health Reports A03-12-I, IPEC Fire Protection Program, Sep.-Oct. 2003 Assessment Report No. 01-AR-01-PG, Indian Point Unit 1&2 Nuclear Quality Assurance &

Oversight Independent Oversight Program; Functional Area: Engineering Programs - Fire Protection, First Quarter 2001 Assessment Report No. 02-AR-20-PG, Fire Protection Program Biennial Assessment, June 2002 ENN-IT-104, Software Quality Assurance Program, Rev. 3 IPEC Focused Self-Assessment Report, Fire Protection & Safe Shutdown Capability for IP2, March 1, 2004 Self Assessment Report, Fire Protection, Rev. 0 IPEC U2 Fire Protection System Health Report - Fourth Quarter 2003 IPEC U2 Fire Protection System Health Report - Third Quarter 2003 QA Surveillance Report, Fire Protection - Fire Drill, January 16, 2002 SE-SQ-12.401, Quality Assurance Document, Pre-Fire Plans, Rev. 6 Attachment

A-6 PI&Ds A227551-60, Fire Protection System Diagram Details, Rev. 60 A227552-40, Fire Protection System Diagram Details, Rev. 40 A227553-49, Fire Protection System Diagram Details, Rev. 30 A227554-24, Fire Protection System Diagram Details, Rev. 24 9321-F-4006-70, Yard Fire Protection Piping, Rev. 70 Fire Protection Drawings 160732-7, IP1 Conduit Layout Smoke Detection System, Rev. 7 A201034, Diesel Generator Building Fire Protection-Spray System Plans and Sections A201078, Electrical Tunnel Fire Protection Spray System Plans & Sections, Rev. 2 A205747-13, Control Building Fire Detection System Control Room, EL. 53-0", Rev.13 A214524-3, IP2 Control Building Elevations & Details, Rev. 3 A214524-6, IP2 Control Building Sections & Details, Rev. 6 B230980, IP2 PAB CURB & Fire Door To Diesel Generator, Rev. 2 EK-8351, Fire Detection System Layout for IP2, Rev. 13 9321-F-3062-35, IP2 Cable Trays, Bus Duct and Conduits In Electrical Tunnel, Rev. 35 Drawings A138040-54, Unit #1 One Line Diagram 13.8KV & 440V System, Rev. 54 A138458-64, IP1 Cable Troughs & Conduit Under Elev. 53-C Co. 6-11, F-M Plan, Rev. 64 A138522-28, IP1 Cable Troughs & Conduit Under Elev. 72'-0" Co. 6-11, F-M Plan, Rev. 28 A138752-15, IP2 Schematic Arrangement - Control & Power Cables - Turbine Room Under Elev. 32'-0", Rev. 13 A141537-36, IP1 Nuclear Service Building Installation Of PWR & Cont. Cond. Elev.

53'-0" & 62'-0" - Plan, Rev. 36 A181899-22, IP2 Lighting & Electrical Heaters Control Passageways, Rev. 22 A206646-19, IP2 Conduit Layout Control Building Elevation 15'-0" Plan, Rev. 19 A206844-09, IP2 Conduit Tray Connection Schematic Turbo Gen Bldg Heater Bay EL 15'-0" Sh 1, Rev. 6 A206845-15, IP2 Conduit And Tray Connection Schematic Turbo Gen Bldg & Heater Bay EL. 15'-0" Sh #2, Rev. 15 A206853-16, IP2 Conduit & Tray Connection Schematic Primary Auxiliary Building EL.

68'-0",Rev. 16 A206854-21, IP2 Conduit & Tray Connection Schematic Primary Auxiliary Building EL.

80'-0", Sheet 1, Rev. 21 A206859-15, IP2 Conduit & Tray Conn. Schematic Control Bldg. EL. 15'-0" Sh.1, Rev.

Attachment

A-7 A206860-8, IP2 Conduit & Tray Conn. Schematic Control Bldg. EL.15-0" SH.2, Rev.

A206861-12, IP2 Tray Schematic Control Bldg. EL.33-0" SH.3, Rev. 12 A206862-16, IP2 Tray Connection Schem. Control Building EL.33-0" Sheet 4, Rev 16 A206869-6, IP2 Conduit & Tray Conn. Schematic-Electrical Tunnel To Penetration, SH.1, Rev. 6 A206870-15, IP2 Conduit & Tray Conn. Schematic Electrical Tunnel To Penetration, Sh.2, Rev. 15 A206861-12, IP2 Tray Schematic, Control Bldg. EL. 33-0" Sh.3, Rev. 12 A206862-16, IP2 Tray Connection Schematic Control Building EL.33-0", Sh. 4, Rev.

A206888-11, IP2 Conduit Connection Schem. For Intake Well Structure Sh. 1, Rev. 11 A206889-11, IP2 Conduit Connection Schem. For Intake Well Structure Sh. 2, Rev. 11 A206976-17, Static Inverter #23, Bypass Sw. #23, Battery Charger #23, 125 VDC Power Panel #23, Battery #23, An Associated Brkrs, Rev. 17 A208064-5, Level & Pressure Instrument Installation For Stm. Generator &

Pressurizer - Arrgt Piping - Instrumentation, Rev. 5 A208088-39, One Line Diag. Of 480 Vac. Swgrs. 21, 22, Bus 2A, 3A, 5A & 6A, Rev. 39 A208168-53, Flow Diagram Chemical & Volume Control System, Rev. 53 A208241-23, Single Line Dia. Of 480 VAC MCC 28A And 211, Rev. 23 A208377-09, Main One Line Diagram, Rev. 9 A208426-3, IP2 D/C Of Ckt Changes For Letdown Line Iso. Valves 201 & 202, Rev. 3 A208500-43, One Line Diag For 480V AC: MCC 26AA And MCC 26BB & 120V AC Dist. Panel 1 & 2, Rev. 43 A208501-35, One Line Diagram 125 VDC For Dist Panels 21, 21A, 21B, 22 And 22A, Rev. 35 A208502-56, Single Line Diagram 118 VAC. 1 Phase Inst. Busses No. 21, 22, 23 And 24, Rev. 56 A208503-30, Schem. Dia Of 118 VAC Inst Busses 21A, 22A, 23A And 24A (Located In CCR), Rev. 30 A208507-35, One Line Diagram 480 VAC MCCs 28 & 210, Rev. 35 A208533-08, Single Line Diagram 125V.D.C. Distr Panels 21AA, 22AA, 23AA & 24AA.

A208826-17, IP2 Wiring Diag. Safety Control Panel SN (JB9), Rev. 17 A209189-19, IP2 Conduits Schematic Control Bldg. EL.33'-0" Sh.3A, Rev. 19 A209190-17, IP2 Conduit Conn. Schem. Control Build. EL.33'-0" Sheet 4A, Rev. 17 A209561-5, Steam Generator & Pressurizer Level & Pressure Instrumentation Arrangement Outside Containment, Rev. 5 A209762-65, Flow Diag. Service Wtr Sys Nuclear Stm Supply Plant Sh. 2 of 2, Rev. 65 A209812-46, General Arrangement Of Equipment - Control Rooms Units 1 & 2 -

UFSAR Figure No. 1.2-7 (Sht. 1), Rev. 46 A227781-75, Flow Diagram Auxiliary Coolant System, Rev. 75 A231592-18, 6900 Volt AC One Line Diagram, Rev. 18 A235296-60, Flow Diagram Safety Injection System, Rev. 60 A244016-19, One Line 440 VAC Swgr. Unit-Substs 11RW1, 12RW3, 12FD3, MCCs 10M, 10N, 10Z & 10X, Rev. 19 A249955-17, One Line Diagram 480V AC MCC 29 & 29A, Rev. 17 Attachment

A-8 A249956-15, One Line Diagram 480V MCC 24 &24A, Rev. 15 A250907-21, Electrical Distribution And Transmission System, Rev. 21 B217913-8, Schem. Dia. Of Conns For Bkr 11, IM & IB - S/S 12FD3 (IP #1), Rev. 8 B248513-10, Single Line Diagram 480V MCC 26C & CCR Vent. Dist. Panel 21, Rev.

IP2-S-000231-04, One-Line Schem. For EDG Building Vent. Dist. Panels #1 & #2, Rev. 4 400400-00, Fire Area/Zone Arrangement Plans at El. 15'-0" and 18'-6", Rev. 0 400401-00, Fire Area/Zone Arrangement Plans at El. 32'-6", 33'-0", 36'-9", 42'-0", and 68'-0", Rev. 0 400402-00, Fire Area/Zone Arrangement Plans at El. 33'-0", 33'-9", 46'-0", 53'-0",

58'-0", Rev. 0 400403-00, Fire Area/Zone Arrangement Plans at El. 46'-0", 53'-0", 64'-0", 68'-0",

72'-0", 80'-0", Rev. 0 400404-00, Fire Area/Zone Arrangement Plans at El. 53'-0", 65'-0", 68'-0", 72'-0",

77'-0", 80'-0", Rev. 0 400405-00, Fire Area/Zone Arrangement Plans at El. 53'-0", 80'-0", 91'-8", 95'-0",

98'-0", Rev. 0 610F970-10, Supervisory Panel SNR Wiring Diagram, Rev. 10 9321-F-2017-83, Flow Diagram Main Steam, Rev. 83 9321-2018-136, Flow Diagram Condensate & Boiler Feed Pump Suction, Rev. 136 9321-F-2019-110, Flow Diagram Boiler Feedwater, Rev. 110 9321-F-2720-81, Flow Diagram Auxiliary Coolant System, Rev. 81 9321-F-2722-107, Flow Diagram Service Water System Nuclear Steam Supply Plant Sheet 1 Of 2, Rev. 107 9321-F-2735-135, Flow Diagram Safety Injection, Rev. 135 9321-F-2736-119, Flow Diagram Chemical & Volume Control System, Rev. 119 9321-F-2738-113, Flow Diagram Reactor Coolant System, Rev. 113 9321-F-3004-82, One Line Dia. 480V Motor Control Centers 21, 22, 23, 25 & 25A, Rev. 82 9321-F-3005-104, One Line Diagram 480V Motor Control Center 27 & 27A, Rev. 104 9321-F-3006-92, Single Line Diagram 480V MCC 26A And 26B, Rev. 92 9321-F-3008-83, Single Line Diagram D.C. Power Panels 21, 22, 23, And 24, Rev. 83 9321-F-3080-58, Conduit Layout Fan Room & Fuel Storage Bldg., Rev. 58 9321-F-3082-75, Conduit Layout Primary Auxiliary Building - Sheet 2, Rev. 75 9321-F-3204-69, 125 VDC Power Panels 21 & 22, 120 VAC Dist. Panels 21 & 22, Rev. 69 9321-F-3207-38, Wiring Diagram 125V. DC Distribution Panels 21 & 22 & 21A, Rev. 38 9321-F-3226-34, Wiring Diag. Of Superv Cont. Pnl SB1 Sheet 1 & Sheet #1 of 2, Rev. 34 9321-F-3237-32, Wiring Diag. Superv. Control Pnl SN, Rev. 32 9321-F-3289-26, Wiring Diagram Protection Logic Rack E-8, Rev. 26 9321-F-3301-37, Wiring Diagram Protection Logic Rack F-7, Rev. 37 9321-F-3307-26, Conduit & Cable Tray Schematic For M.C.C. 26A, 26B, 27 & Lighting SWGR. S.E. Corner Of P.A.B. EL. 80'-0", Rev. 26 Engineering Evaluations and Penetration Seal Fire Tests 2003-05, Appendix R Review & Combustible Loading Summary 2003-09, Appendix R Review & Combustible Loading Summary Attachment

A-9 2003-10, Appendix R Review & Combustible Loading Summary 2003-38, Appendix R Review & Combustible Loading Summary CTP 1001A, Three Hour Fire Qualification Test 10" and 6 Depth Silicon RTV Foam For Electrical & Mechanical Penetration Seals, July 25, 1980 ICC0483032, Fire & Hose -Stream Tests for Penetration Seal Systems, May 1983 Calculations CN-PO-04-2, Indian Point Unit 2 Appendix R Loss of All AC Power Recovery Analysis Using Treat 2.02, Rev. 1 ECX-00020-00, Short Circuit Evaluation Of Gaps In Cable Tray Barriers And Dividers, Rev. 0 EGP-S80-009-2, Metal Dividers For Heavy Power Cable Trays EGP-00036-00, Evaluation Of Short Circuit Effects On Heavy Power Cables For SIP22 and CP21, Rev. 0 EI-2028, Protective Settings & Coordination Criteria, Indian Point Unit 2, Calculation SGX-00058-00, Attachment 4, Revision 1 FCX-00035-00, Operability Analysis of Stairwell #4 Fire Protection Piping, April 25, 1997 FEX-00141, IP2 Amptector Setting Verification, Sensor and Tolerances (480v Switchgear), Rev. 0 PGI-00403, Fire Barrier Penetration Seal Database, April 6, 2001 PGI-00460, Evaluation of The High Pressure fire Water Supply System, Rev. 0 PGI-00581, Fire Protection 86-10 Evaluation for EDG Structural Steel Fire Resistant Coating, Rev. 0 PGI-04008, Required Volume of the Unit 2 Diesel Driven Fire Pump Fuel Oil Storage Tank, Rev. 0 Fire Protection Specifications 9321-01-44-1, Yard Storm Drainage and Yard and Building Standpipe Fire Protection System, Rev. 1 9321-01-44-2, Floor, Equipment and Roof Drains, Rev. 1 9321-01-70-1, Water Spray & Foam Fire Protection for IP2, February 28, 1969 9321-01-70-4, Fire Protection Spray System For Cable Tunnel for IP2, November 5, 1970 9321-01-135-4, Fire Detection System for IP2, November 18, 1968 Hot Work Permits W.O. IP2-02-00452, 12 Coldwater Vacuum Pump W.O. IP2-02-00540, Chemical feed For AFW To 21 Steam Generator Drain Valve W.O. IP2-02-02177, Aux. Steam Supply Stop R4D4 W.O. IP2-03-19925, 68' CCW 761A Check Valve W.O. IP2-03-24046, 15' THB W.O. IP2-03-24661, TSC Emergency Diesel Generator Attachment

A-10 Condition Reports CR-IP2 1997-03731 1999-04426 1999-07993 1999-00192 1995-04759 1997-04136 1997-05174 1998-10139 1998-10141 1998-10143 1999-00192 1999-03202 1999-04426 1999-04922 1999-05145 1999-06568 1999-07993 1999-09451 2000-03798 2000-03822 2000-07367 2000-07398 2001-01570 2001-02350 2001-02351 2001-02352 2001-02353 2001-02354 2001-02355 2001-02357 2001-02358 2001-02359 2001-02360 2001-02363 2001-02364 2001-02366 2001-02367 2001-02370 2001-02371 2001-02372 2001-02373 2001-02374 2001-02375 2001-02376 2001-02377 2001-02378 2001-02379 2001-02381 2001-03417 2001-03730 2001-08973 2001-09976 2002-01859 2002-09782 2003-00765 2003-00867 2003-01508 2003-02428 2003-04387 2003-05837 2003-06341 2003-06834 2003-06848 2003-07227 2003-07541 2003-07605 2003-07608 2004-00320 2004-00460 2004-00509 2004-00568 2004-00609 2004-01002 2004-01031 2004-01100 2004-01104 2004-01247 2004-01257 2004-01258 2004-01259 2004-01260 2004-01266 2004-01267 2004-01269 2004-01299 2004-0130 2004-01310 2004-01312 2004-01338 2004-01445 Work Orders (W.O.)

NP-99-12365 NP-00-14398 NP-00-14474 NP-00-14501 NP-00-14969 NP-01-22486 IP2-01-23830 IP2-01-25129 IP2-02-31353 IP2-02-32719 IP2-02-32720 IP2-02-33445 IP2-03-05682 IP2-03-10775 IP2-03-14039 IP2-03-16327 IP3-03-10139 Modifications ER IP2-03-21959, 21 Charging Pump Cable Separation, January 13, 2004 Miscellaneous Documents Active Surveillance Test List, February 5, 2004 Design Basis Improvement Project PI-1, Fire Protection Improvement Plan, Rev. 1.

Design Basis Document For Electrical Separation, IP2-DBD-222, Revision 1, 12/17/03 Emergency Systems, Inc., Big Beam Emergency Lighting Equipment - Vendor Manual Fire Protection Impairment Summary, February 27, 2004 Attachment

A-11 IP2 Maintenance Rule Basis Document for FP, Rev. 0 NFPA 20-1978, Centrifugal Fire Pumps Penetration Seal Index, 11/39A-W001 and 11/39A-W021 PGI-00403-00, Fire Barrier Penetration Seal Database, Rev. 0 LIST OF ACRONYMS USED ASSS Alternate Safe Shutdown System CA Corrective Action CCW Component Cooling Water CFR Code of Federal Regulations CR Condition Report DRS Division of Reactor Safety FA Fire Area FPPP Fire Protection Program Plan FZ Fire Zone GL Generic Letter gpm Gallons Per Minute ICA Implement Corrective Action IP Inspection Procedure IPE Individual Plant Examination IPEEE Individual Plant Examination of External Events IR Inspection Report LER Licensee Event Report MOV Motor Operated Valve NCV Non-Cited Violation NFPA National Fire Protection Association NRC Nuclear Regulatory Commission PAR Publicly Available Record P&ID Piping and Instrumentation Drawing PORV Power Operated Relief Valve PWR Pressurized Water Reactor QA Quality Assurance RCP Reactor Coolant Pump RCS Reactor Coolant System SCBA Self-Contained Breathing Apparatus SDP Significance Determination Process SER Safety Evaluation Report SI Safety Injection UFSAR Updated Final Safety Analysis Report URI Unresolved Item V Volt WOG Westinghouse Owners Group Attachment