IR 05000293/2009007
| ML093020045 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 10/28/2009 |
| From: | Rogge J Engineering Region 1 Branch 3 |
| To: | Bronson K Entergy Nuclear Operations |
| References | |
| IR-09-007 | |
| Download: ML093020045 (26) | |
Text
ber 28, 2009
SUBJECT:
PILGRIM NUCLEAR POWER STATION - NRC TRIENNIAL FIRE PROTECTION INSPECTION REPORT 05000293/2009007
Dear Mr. Bronson:
On October 1, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Pilgrim Nuclear Power Station. The enclosed inspection report documents the inspection results which were discussed at an exit meeting on October 1, 2009, with you and other members of your staff.
The 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.
The report documents one NRC-identified finding of very low safety significance (Green) that was determined to involve a violation of NRC requirements. Additionally, a licensee-identified violation which was determined to be of very low safety significance is listed in this report.
However, because of the very low safety significance and because they were entered into your corrective action program, the NRC is treating the findings as non-cited violations (NCVs)
consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest any NCV, you should provide a response within 30 days of the date of this report, with the basis for your denial, to the Nuclear Regulatory Commission, Region I; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Pilgrim Nuclear Power Station. In addition, if you disagree with the characterization of any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region I, and the Resident Inspector at the Pilgrim Nuclear Power Station. The information you provide will be considered in accordance with Inspection Manual Chapter 0305. In accordance with Title 10 of the Code of Federal Regulations Part 2.390 of the NRCs Rules of Practice, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the 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).
Sincerely,
/RA/
John F. Rogge, Chief Engineering Branch 3 Division of Reactor Safety Docket No. 50-293 License No. DPR-35 Enclosure: Inspection Report No. 05000293/2009007 w/Attachment: Supplemental Information cc w/encl: Distribution via ListServ
SUMMARY OF FINDINGS
IR 05000293/2009007; 09/14/2009-10/01/2009; Entergy Nuclear Operations, Inc.; Pilgrim
Nuclear Power Station; Triennial Fire Protection Team Inspection.
The report covered a two-week triennial fire protection team inspection by four Region I specialist inspectors. One Green NCV was identified. The significance of most findings is indicated by their color using Inspection Manual Chapter 609, Significance Determination Process, (SDP) and the cross-cutting aspect was evaluated using IMC 0305, Operating Reactor Assessment Program. Findings for which the SDP does not apply may be Green or assigned a severity level after NRC management review. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.
Cornerstone: Mitigating Systems
- Green.
The inspectors identified that Entergy did not ensure that plant procedures were adequate to prevent a spurious actuation from affecting the ability to provide a source of make-up water to the reactor vessel within 20 minutes following the evacuation of the control room during a fire as specified in procedure 2.4.143, Shutdown From Outside the Control Room,
Revision 40. The finding was determined to be of very low safety significance (Green) and a NCV of the Pilgrim Nuclear Power Station Technical Specification 5.4.1.d, Procedures. Entergy entered the issue into the corrective action program and planned to implement changes to the procedure to resolve the issue. Entergy also reviewed completed reactor core isolation cooling (RCIC) and high pressure coolant injection (HPCI) system startup job performance measures (JPMs) and performed procedure walkthroughs to assess the time needed to attempt a RCIC start and then transfer to, and start HPCI to confirm these actions could be taken in within the time necessary to prevent lowering vessel level to that of the top of active fuel.
The inspectors determined that this finding was more than minor because it was associated with the procedure quality attribute of the mitigating system cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, during a fire in the control room or cable spreading room there are four valves in each of the RCIC and HPCI systems that could spuriously close due to fire damage to cabling. Procedure 2.4.143 does not ensure that the associated motor control center circuit breakers are opened (to prevent spurious closure) and that the valves are in the correct position prior to starting one of the systems to provide make-up to the reactor vessel. Unidentified spurious closure of the valves during or after startup of the systems could disable the system and delay establishment of reactor vessel makeup. The inspectors assessed this finding in accordance with NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process. This finding affected safe shutdown capabilities and screened to very low safety significance (Green) in Phase 1 of the SDP because it was assigned a low degradation rating. A low degradation rating was assigned because it was determined to be a minor procedure issue that could be compensated for by operator experience and familiarity. No cross-cutting aspect was assigned because the inspectors concluded this issue was not indicative of current licensee performance. (Section 1R05.01)ii
Other Findings
A violation of very low safety significance was identified by Entergy and has been reviewed by the inspectors. Corrective actions taken by Entergy have been entered into their corrective action program. The NCV and corrective action tracking number are listed in Section 4OA7 of this report.
iii
REPORT DETAILS
Background This report presents the results of a triennial fire protection inspection conducted in accordance with NRC Inspection Procedure (IP) 71111.05T, Fire Protection. The objective of the inspection was to assess whether Entergy Nuclear Operations, Inc. has implemented an adequate fire protection program and that post-fire safe shutdown capabilities have been established and are being properly maintained at the Pilgrim Nuclear Power Station (PNPS).
The following fire areas (FAs) and fire zones (FZs) were selected for detailed review based on risk insights from the PNPS Individual Plant Examination (IPE)/Individual Plant Examination of External Events (IPEEE):
- FA 3.1
- FA 1.10 (FZ 2.1 and 2.4)
- FA 1.10 (FZ 4.1)
- FA 1.10 (FZ 1.28)
Inspection of these areas/zones fulfills the inspection procedure requirement to inspect a minimum of three samples.
The inspectors evaluated the licensees fire protection program (FPP) against applicable requirements which included plant Technical Specifications, Operating License Condition 3.F, NRC Safety Evaluations, 10 CFR 50.48, and 10 CFR 50, Appendix R. The inspectors also reviewed related documents that included the Final Safety Analysis Report (FSAR), Section 10.8, the fire protection program (FPP), and the post-fire safe shutdown analysis.
Specific documents reviewed by the inspectors are listed in the attachment.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R05 Fire Protection (IP 71111.05T)
.01 Post-Fire Safe Shutdown from Outside Main Control Room (Alternative Shutdown) and
Normal Shutdown
a. Inspection Scope
Methodology The inspectors reviewed the safe shutdown analysis, operating procedures, piping and instrumentation drawings (P&IDs), electrical drawings, the FSAR and other supporting documents to verify that hot and cold shutdown could be achieved and maintained for fires that rely on shutdown from outside the control room. This review included verification that shutdown from outside the control room could be performed both with and without the availability of offsite power. Plant walkdowns were also performed to verify that the plant configuration was consistent with that described in the safe shutdown and fire hazards analyses. These inspection activities focused on ensuring the adequacy of systems selected for reactivity control, reactor coolant makeup, reactor decay heat removal, process monitoring instrumentation, and support systems functions.
The inspectors verified that the systems and components credited for use during this shutdown method would remain free from fire damage. The inspectors verified that the transfer of control from the control room to the alternative shutdown location(s) would not be affected by fire-induced circuit faults (e.g., by the provision of separate fuses and power supplies for alternative shutdown control circuits).
Similarly, for fire areas that utilize shutdown from the control room, the inspectors also verified that the shutdown methodology properly identified the components and systems necessary to achieve and maintain safe shutdown conditions.
Operational Implementation The inspectors verified that the training program for licensed and non-licensed operators included alternative shutdown capability. The inspectors also verified that personnel required for safe shutdown using the normal or alternative shutdown systems and procedures are trained and available onsite at all times, exclusive of those assigned as fire brigade members.
The inspectors reviewed the adequacy of procedures utilized for post-fire shutdown and performed an independent walk through of procedure steps to ensure the implementation and human factors adequacy of the procedures. The inspectors also verified that the operators could be reasonably expected to perform specific actions within the time required to maintain plant parameters within specified limits. Time critical actions, which were verified, included restoration of alternating current (AC) electrical power, establishing the remote shutdown and local shutdown panels, establishing reactor coolant makeup, and establishing decay heat removal.
Specific procedures reviewed for alternative shutdown, including shutdown from outside the control room included the following:
- 2.4.143, Shutdown From Outside The Control Room, Rev. 40
- 2.4.143.2, Shutdown With a Fire In Reactor Building West (Fire Area 1.10),
Rev. 16
- 5.5.1, General Fire Procedure, Rev. 25 The inspectors reviewed manual actions to ensure that they had been properly reviewed and approved and that the actions could be implemented in accordance with plant procedures in the time necessary to support the safe shutdown method for each fire area. The inspectors also reviewed the periodic testing of the alternative shutdown transfer capability and instrumentation and control functions to ensure the tests are adequate to ensure the functionality of the alternative shutdown capability.
b. Findings
===.1
Introduction.
The inspectors identified that Entergy did not ensure that plant procedures===
were adequate to prevent a spurious valve actuation from affecting the ability to provide a source of make-up water to the reactor vessel within 20 minutes following the evacuation of the control room during a fire as specified in procedure 2.4.143, Shutdown From Outside the Control Room, Revision 40. The finding was determined to be of very low safety significance (Green) and a NCV of the Pilgrim Nuclear Power Station technical specification 5.4.1.d, Procedures.
Description.
PNPS procedure 2.4.143, Shutdown From Outside Control Room, provides instructions to plant operators for control of the plant from remote shutdown panels in the event of a control room fire or cable spreading room fire that requires evacuation of the control room. The procedure directs the starting of the RCIC or HPCI system to inject into the reactor pressure vessel (RPV) within 20 minutes of abandoning the control room. The procedure also includes directions to perform electrical circuit breaker alignments to prevent spurious operation of components that could occur as a result of fire damage to control cables.
There are four valves in both the RCIC and HPCI systems that have the potential for spurious operation. The affected valves include two isolation valves in each of the turbine steam supply line for each turbine, the pump discharge valve for each system, and the pump suction valve from the condensate storage tank for each system. Closure of any of these valves would prevent operation of the system and delay the addition of water to the RPV. The inspectors also noted that closure of the suction valve on either of the pumps could result in damage to the pump if operators did not promptly identify the closure and take the necessary actions to shut down the system. Again, shutting down the system, correcting the valve position and restarting the system would delay injection to the RPV.
The inspectors noted that the procedure included precautions to alert the operators to the potential for spurious valve closures. However, the inspectors found that the procedure step sequence instructs the operator to start the RCIC or HPCI system (per procedure Attachments B or A respectively) prior to performing the electrical alignments per Attachment F of the procedure. With this sequence, the systems are vulnerable to spurious valve operations and resultant delays in injection to the RPV.
Analysis.
The inspectors determined that the failure to provide an appropriate procedure step sequence that addresses the potential adverse impact of spurious valve operation was a performance deficiency that was within the ability of the licensee to correct. The inspectors determined that this finding was more than minor because it was associated with the procedure quality attribute of the mitigating system cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage).
The inspectors assessed this finding in accordance with NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process. This finding affected safe shutdown capabilities and screened to very low safety significance (Green) in Phase 1 of the SDP because it was assigned a low degradation rating. A low degradation rating was assigned because it was determined to be a minor procedure issue that could be compensated for by operator experience and familiarity.
The inspectors noted that the procedure specifically requires that the RCIC and HPCI system operations be performed by licensed reactor operators. The inspectors also found that operator training program included specific job performance measures (JPMs)that test the operator on startup and operation of the RCIC and HPCI systems from the remote shutdown panels. Additionally, JPM-217-08, RCIC Start From The Alternate Shutdown Panels, Rev. 2, includes training associated with the potential spurious closure of valves. Additional factors considered by the inspectors were that both HPCI and RCIC can be operated remotely thereby providing a second method of injection to the vessel should one system be damaged by spurious closure of a pump suction valve.
The inspectors also noted the 20 minute time limit imposed by the procedure is more restrictive the engineering calculation conclusions which supported a time limit of approximately 27 minutes. Furthermore, if control rod drive system injection was not lost due to a loss of offsite power, or is promptly restored following a loss of offsite power, the time limit to initiate RCIC or HPCI injection would be lengthened. The licensee also performed procedure walkthroughs to assess the time required for an operator to attempt a start of RCIC followed by a transfer to, and startup of HPCI if the RCIC start was unsuccessful. While these times exceeded the 20 minute limit of the plant procedure, the time was estimated to be approximately 24 minutes and thereby establish RPV injection prior to vessel level dropping to the top of active fuel region.
No cross-cutting aspect was assigned because the inspectors concluded this issue was not indicative of current licensee performance. (Section 1R05.01)
Enforcement.
PNPS Technical Specification 5.4.1.d requires that written procedures be established, implemented and maintained covering the implementation of the fire protection program. Contrary to this requirement, Entergy failed to establish adequate procedures to prevent loss of the RCIC or HPCI systems due to spurious valve operations caused by the fire. As a result, the 20 minute time limit for establishing makeup to the RPV, as specified in the procedure, may not have been met. Because this violation was of very low safety significance and it was entered into the licensees corrective action program (CRs 2009-4205 and 2009-4240), this violation is being treated as an NCV consistent with Section VI.A.1 of the NRC Enforcement Policy. (NCV 05000293/2009007-01, Failure to establish adequate procedures to prevent adverse impact due to spurious valve closure caused by fire damage.)
.02 Protection of Safe Shutdown Capabilities
a. Inspection Scope
The inspectors reviewed the FHA, safe shutdown analyses, and supporting drawings and documentation to verify that safe shutdown capabilities were properly protected.
The inspectors ensured that separation requirements of Section III.G of 10 CFR 50, Appendix R were maintained for the credited safe shutdown equipment and their supporting power, control, and instrumentation cables. This review included an assessment of the adequacy of the selected systems for reactivity control, reactor coolant makeup, reactor heat removal, process monitoring, and associated support system functions.
The inspectors reviewed the licensees procedures and programs for the control of ignition sources and transient combustibles to assess their effectiveness in preventing fires and in controlling combustible loading within limits established in the FHA. A sample of hot work and transient combustible control permits were also reviewed. The inspectors performed plant walkdowns to verify that protective features were being properly maintained and administrative controls were being implemented.
b. Findings
No findings of significance were identified.
.03 Passive Fire Protection
a. Inspection Scope
The inspectors walked down accessible portions of the selected fire areas to observe material condition and the adequacy of design of fire area boundaries (including walls, fire doors and fire dampers), and electrical raceway fire barriers to ensure they were appropriate for the fire hazards in the area.
The inspectors reviewed installation/repair and qualification records for a sample of penetration seals to ensure the fill material was of the appropriate fire rating and that the installation met the engineering design. The inspectors also reviewed similar records for the fire protection wraps to ensure the material was of an appropriate fire rating and that the installation met the engineering design.
b. Findings
No findings of significance were identified.
.04 Active Fire Protection
a. Inspection Scope
The inspectors reviewed the design, maintenance, testing, and operation of the fire detection and suppression systems in the selected plant fire areas. This included verification that the manual and automatic detection and suppression systems were installed, tested, and maintained in accordance with the National Fire Protection Association (NFPA) code of record, or as NRC approved exemptions, and that each suppression system would control and/or extinguish fires associated with the hazards in the selected areas. A review of the design capability of the suppression agent delivery systems were verified to meet the code requirements for the hazards involved. The inspectors also performed a walkdown of accessible portions of the detection and suppression systems in the selected areas as well as a walkdown of major system support equipment in other areas (e.g. fire pumps, Halon and/or carbon dioxide (CO2)storage tanks and supply system) to assess the material condition of the systems and components.
The inspectors reviewed electric and diesel fire pump flow and pressure tests to ensure that the pumps were meeting their design requirements. The inspectors also reviewed the fire main loop flow tests to ensure that the flow distribution circuits were able to meet the design requirements.
The inspectors assessed the fire brigade capabilities by reviewing training, qualification, and drill critique records. The inspectors also reviewed pre-fire plans and smoke removal plans 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 post-fire safe shutdown capability. In addition, the inspectors inspected the fire brigade equipment (including smoke removal equipment) to determine operational readiness for fire fighting.
b. Findings
No findings of significance were identified.
.05 Protection from Damage from Fire Suppression Activities
a. Inspection Scope
The inspectors performed document reviews and plant walkdowns to verify that redundant trains of systems required for hot shutdown are not subject to damage from fire suppression activities or from the rupture or inadvertent operation of fire suppression systems. Specifically, the inspectors verified that:
- A fire in one of the selected fire areas would not directly, through production of smoke, heat or hot gases, cause activation of suppression systems that could potentially damage all redundant safe shutdown trains.
- A fire in one of the selected fire areas (or the inadvertent actuation or rupture of a fire suppression system) would not directly cause damage to all redundant trains (e.g. sprinkler caused flooding of other than the locally affected train).
- Adequate drainage is provided in areas protected by water suppression systems.
b. Findings
No findings of significance were identified.
.06 Alternative Shutdown Capability
a. Inspection Scope
Alternative shutdown capability is discussed in section 1R05.01 of this report.
.07 Circuit Analysis
a. Inspection Scope
The inspectors verified that the licensee performed a post-fire safe shutdown analysis for the selected fire areas and the analysis appropriately identified the structures, systems, and components important to achieving and maintaining safe shutdown. Additionally, the inspectors verified that the licensees analysis ensured that necessary electrical circuits were properly protected and that circuits that could adversely impact safe shutdown due to hot shorts, shorts to ground, or other failures were identified, evaluated, and dispositioned to ensure spurious actuations would not prevent safe shutdown.
The review considered fire and cable attributes, potential undesirable consequences and common power supply/bus concerns. Specific items included the credibility of the fire threat, cable insulation attributes, cable failure modes, and actuations resulting in flow diversion or loss of coolant events.
The inspectors also reviewed cable raceway drawings for a sample of components required for post-fire safe shutdown to verify that cables were routed as described in the cable routing matrices.
Cable failure modes were reviewed for the following components:
- MO-1301-16, RCIC Steam Supply Valve
- MO-1301-22, RCIC Pump Suction Valve
- MO-1301-48, RCIC Pump Discharge Valve
- MO-2301-4, HPCI Steam Supply Valve
- MO-2301-6, HPCI Pump Suction Valve
- MO-2301-9, HPCI Pump Discharge Valve The inspectors reviewed 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 inspectors confirmed that coordination studies had addressed multiple faults due to fire.
b. Findings
No findings of significance were identified.
.08 Communications
a. Inspection Scope
The inspectors reviewed safe shutdown procedures, the safe shutdown analysis, and associated documents to verify an adequate method of communications would be available to plant operators following a fire. During this review the inspectors considered the effects of ambient noise levels, clarity of reception, reliability, and coverage patterns.
The inspectors also inspected the designated emergency storage lockers to verify the availability of portable radios for the fire brigade and for plant operators. The inspectors also verified that communications equipment such as repeaters and transmitters would not be affected by a fire.
b. Findings
No findings of significance were identified.
.09 Emergency Lighting
a. Inspection Scope
The inspectors observed the placement and coverage area 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 and/or instrumentation monitoring for post-fire safe shutdown. The inspectors also verified that the battery power supplies were rated for at least an eight-hour capacity. Preventive maintenance procedures, the vendor manual, completed surveillance tests, and battery replacement practices were also reviewed to verify that the emergency lighting was being maintained in a manner that would ensure reliable operation.
b. Findings
No findings of significance were identified.
.10 Cold Shutdown Repairs
a. Inspection Scope
The inspectors did not review cold shutdown repair aspects of the 71111.05T procedure because PNPS does not credit repairs to achieve cold shutdown.
b. Findings
No findings of significance were identified.
.11 Compensatory Measures
a. Inspection Scope
The inspectors verified that compensatory measures were in place for out-of-service, degraded or inoperable fire protection and post-fire safe shutdown equipment, systems, or features (e.g. detection and suppression systems and equipment, passive fire barriers, or pumps, valves or electrical devices providing safe shutdown functions or capabilities). The inspectors also verified that the short term compensatory measures compensated for the degraded function or feature until appropriate corrective action could be taken and that the licensee was effective in returning the equipment to service in a reasonable period of time.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
[OA]
4OA2 Identification and Resolution of Problems
.01 Corrective Actions for Fire Protection Deficiencies
a. Inspection Scope
The inspectors verified that the licensee was identifying fire protection and post-fire safe shutdown issues at an appropriate threshold and entering them into the corrective action program. The inspectors also reviewed a sample of selected issues to verify that the licensee had taken or planned appropriate corrective actions.
b. Findings
No findings of significance were identified.
4OA6 Meetings, Including Exit
Exit Meeting Summary
The inspectors presented their preliminary inspection results to Mr. K. Bronson, Vice President PNPS, and other members of the site staff at an exit meeting on October 1, 2009. No proprietary information was included in this inspection report.
4OA7 Licensee-Identified Violations
The following violation of very low safety significance (Green) was identified by Entergy and is a violation of NRC requirements which met the criteria of the NRC Enforcement Policy for disposition as a Non-Cited Violation (NCV).
- License Condition 3.F for Pilgrim Nuclear Power Station (PNPS) states in part that, Entergy shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Report (SER)dated December 21, 1978. The SER, states in part, an analysis of safety related shutdown systems will be conducted to demonstrate that fire-related damage in any fire area will not inhibit the capability to safely shutdown.
Contrary to this requirement, Entergy calculation PS32, Appendix R Safe Shutdown Analysis Report, incorrectly assumed that the reactor head vent valves (SV220-46 and SV220-47) would not be vulnerable to spurious operation (opening) for a fire in the control room or cable spreading room and would remain closed while the reactor was at power. During review of a recent modification, the licensee determined that the head vent valves could spuriously open due to a control room or cable spreading room fire during a potential control room evacuation scenario. The issue was entered into Entergys corrective action program (CR 2009-01376) and corrected by implementation of a modification (EC 14506) to the head vent valve circuits. PS32 was subsequently updated to reflect this change.
The inspectors determined that the finding was of very low safety significance (Green) because appropriate plant procedures were in place to operate systems required for post-fire safe shutdown during a control room evacuation scenario and those systems would be have been able to provide adequate makeup to the reactor vessel to compensate for inventory loss through the head vent piping.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
S. Burke Fire Protection Engineer
F. Clifford Senior Reactor Operator
F. McGinnis Licensing Engineer
- D. Berkland Design Engineer, Electrical
- S. Das Design Engineer, Electrical
E. Hinxman Reactor Operator
- J. House Superintendent, Training
M. Landry System Engineer
C. McMorrow Fire Brigade Training
NRC
- J. Rogge Chief, Engineering Branch 3, Division of Reactor Safety
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
- 05000293/2009007-01 NCV Failure to establish adequate procedures to prevent adverse impact due to spurious valve closure caused by fire damage.