IR 05000293/2007002
| ML071270525 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 05/07/2007 |
| From: | Racquel Powell NRC/RGN-I/DRP/PB5 |
| To: | Bronson K Entergy Nuclear Operations |
| Powell R, RI/DRP/610-337-6967 | |
| References | |
| IR-07-002 | |
| Download: ML071270525 (31) | |
Text
UNITED STATES May 7, 2007
SUBJECT:
PILGRIM NUCLEAR POWER STATION - NRC INTEGRATED INSPECTION REPORT 05000293/2007002
Dear Mr. Bronson:
On March 31, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Pilgrim Nuclear Power Station. The enclosed integrated inspection report documents the inspection findings, which were discussed on April 5, 2007, with you and 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.
This report documents one self-revealing finding of very low safety significance (Green), which involved a violation of NRC requirements. Additionally, two licensee-identified violations which were determined to be of very low safety significance are listed in this report. However, because of the very low safety significance and because they have been entered into your corrective action program, the NRC is treating these findings as non-cited violations (NCVs),
consistent with Section VI.A.1 of the NRC's Enforcement Policy. If you contest any NCV in this report, you should provide a response with the basis for your denial, within 30 days of the date of this inspection report, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington, D.C. 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C.
20555-0001; and the NRC Resident Inspector at Pilgrim.
In accordance with 10 CFR 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/
Raymond J. Powell, Chief Projects Branch 5 Division of Reactor Projects Docket No. 50-293 License No. DPR-35 Enclosure: Inspection Report 05000293/2007002 w/Attachment: Supplemental Information cc w/encl:
G. J. Taylor, Chief Executive Officer, Entergy Operations M. Kansler, President, Entergy Nuclear Operations, Inc.
J. T. Herron, Senior Vice President M. Balduzzi, Senior Vice President, Northeastern Regional Operations C. Schwarz, Vice-President, Operations Support S. J. Bethay, Director, Nuclear Safety Assurance O. Limpias, Vice President, Engineering J. F. McCann, Director, Licensing C. D. Faison, Manager, Licensing R. Patch, Director of Oversight, Entergy Nuclear Operations, Inc.
B. S. Ford, Manager, Licensing, Entergy Nuclear Operations, Inc.
T. C. McCullough, Assistant General Counsel S. Lousteau, Treasury Department, Entergy Services, Inc.
Director, Radiation Control Program, Commonwealth of Massachusetts W. Irwin, Chief, CHP, Radiological Health, Vermont Department of Health The Honorable Therese Murray The Honorable Vincent deMacedo Chairman, Plymouth Board of Selectmen Chairman, Duxbury Board of Selectmen Chairman, Nuclear Matters Committee Plymouth Civil Defense Director D. OConnor, Massachusetts Secretary of Energy Resources J. Miller, Senior Issues Manager Office of the Commissioner, Massachusetts Department of Environmental Protection Office of the Attorney General, Commonwealth of Massachusetts Electric Power Division, Commonwealth of Massachusetts R. Shadis, New England Coalition Staff D. Katz, Citizens Awareness Network Chairman, Citizens Urging Responsible Energy
SUMMARY OF FINDINGS
IR 05000293/2007-002; 01/01/2007-03/31/2007; Pilgrim Nuclear Power Station;
Post-Maintenance Testing.
The report covered a 13-week period of inspection by resident and region-based inspectors.
One Green finding, which was a non-cited violation (NCV), was identified. The significance of most findings is indicated by their color (greater than Green, or Green, White, Yellow, Red)using IMC 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRCs program for overseeing the safe operation of nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.
NRC-Identified and Self-Revealing Findings
- Green.
A Green self-revealing NCV of 10 CFR 50, Appendix B, Criterion XVI,
Corrective Action, was identified for Entergys failure to promptly correct a condition adverse to quality associated with the B emergency diesel generator (EDG). During the post overhaul surveillance of the B EDG on January 25, 2007, the B EDG experienced unexpected load oscillations of approximately 150 kilowatt (kW).
Subsequently, on February 23, 2007, oscillations of greater than 200 kW were seen, which resulted in the shutdown of the B EDG and an entry into a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Technical Specification (TS) Limiting Condition for Operation (LCO). Entergy corrected the kW load oscillations by replacing the mechanical portion of the B EDG governor. The B EDG was declared operable following successful testing. The issue was entered into Entergys corrective action program.
The inspector determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the reliability, availability, and capability of systems that respond to initiating events to prevent undesirable consequences. A Phase 3 SDP evaluation was necessary due to a potential for a greater than green finding as indicated in the site specific pre-solved Phase 2 worksheets. The Phase 3 evaluation concluded that the finding was of very low safety significance (Green). The inspector also determined that this finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, in that Entergy personnel failed to thoroughly evaluate the unexpected kW oscillations. (Section 1R19)
Licensee-Identified Violations
Two violations of very low safety significance, which were identified by the licensee, have been reviewed by the inspector. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. The violations and corrective actions are listed in Section 4OA7 of this report.
iii
REPORT DETAILS
Summary of Plant Status
Pilgrim Nuclear Power Station began the inspection period operating at 100 percent core thermal power. Plant power was reduced to 62 percent on January 26, 2007, to perform power suppression testing to identify and suppress a suspected fuel defect. The plant was returned to full power the next day. The plant began end-of-cycle coast down on January 30, 2007. The plant was shutdown on March 17, 2007, in response to an increase in unidentified drywell leakage. The plant was restarted on March 19, 2007, after repair of a packing leak on reactor water cleanup (RWCU) valve MO-1201-85. The plant was operating at 83 percent power in end-of-cycle coast down at the end of the inspection period.
REACTOR SAFETY
Cornerstones: Initiating Events, Barrier Integrity, Mitigating Systems
1R01 Adverse Weather Protection
a.
Inspection Scope (2 samples)
The inspector reviewed licensee activities to protect plant systems during adverse weather during the periods of January 16-18 and 25-26, 2007 (cold temperatures), and February 14-15, 2007 (coastal storm). The inspector reviewed Entergys actions to mitigate the impact of adverse weather on key plant systems. The review focused on outdoor tanks and on environmental conditions in several buildings, including the intake, emergency diesel generator (EDG), and station blackout diesel generator (SBODG)buildings. The plant systems reviewed included the EDG, the SBODG, the fire water system, the demineralized water supply system, and the salt service water (SSW)system. The references used during this review included station procedures 2.1.37, Coastal Storms; 8.C.40, Cold Weather Surveillance, Attachment 2; and Section 10.9.3 of the Updated Final Safety Analysis Report (UFSAR). The inspector confirmed that Entergy was identifying weather related issues and had entered them into the corrective action program. Additional references used for this review are listed in the attachment to this report. This inspection activity represents two samples for the onset of adverse weather.
b. Findings
No findings of significance were identified.
1R04 Equipment Alignment
.1 Partial System Walkdowns
a.
Inspection Scope (4 samples)
The inspector completed a partial review of the risk significant systems listed below to determine whether the systems were correctly aligned to perform their designated safety functions. The reviews occurred during periods when a redundant train or system was out-of-service for maintenance and/or testing, or following restoration of the system or train from maintenance. The position of key valves, breakers, and control switches required for system operability were verified by field walkdown and/or review of the main control board indications. To ascertain the required system configuration, the inspector reviewed plant procedures, system drawings, the UFSAR, and Technical Specifications (TS). The references used for this review are listed in the attachment to this report.
This inspection activity represents four samples.
- A EDG on January 23, 2007;
- A Residual Heat Removal (RHR) System on January 5, 2007;
- A Core Spray on January 29, 2007; and
- Reactor Core Isolation Cooling (RCIC) System on February 21, 2007.
b. Findings
No findings of significance were identified.
.2 Complete System Walkdowns
a. Inspection Scope
(1 sample)
The inspector performed a full system review of the A SSW Loop, including the C SSW Pump, on February 11, 2007, to verify the system was properly aligned and capable of performing its safety function. To ascertain the required system configuration, the inspector reviewed plant procedures, system drawings, the UFSAR, and the TS. The references used for this review are listed in the attachment to this report. A walkdown of the accessible portions of the system was performed to assess the systems material condition and the following attributes:
- valves were correctly positioned and did not exhibit leakage that would impact the function(s) of any given valve;
- electrical power was available and properly aligned;
- major system components were properly labeled;
- hangers and supports were correctly installed and functional;
- ancillary equipment or debris did not interfere with system performance; and
- valves were locked as required by the locked valve program.
This inspection activity represents one sample.
b. Findings
No findings of significance were identified.
1R05 Fire Protection
.1 Quarterly Fire Protection Inspection
a. Inspection Scope
(12 samples)
The inspector toured selected areas of the plant to observe conditions related to:
- (1) transient combustibles and ignition sources;
- (2) fire detection systems;
- (3) manual firefighting equipment and capability; and
- (4) passive fire protection features. The inspector reviewed the material condition of active and passive fire protection system features, and their operational lineup and readiness. The inspector also reviewed the applicable fire hazard analysis fire zone data sheets. The inspector verified that the licensee addressed fire protection deficiencies in the corrective action program. The references used for this review are listed in the attachment to this report. This inspection activity represents twelve samples.
- UPS [Uninterruptible Power Supply] Diesel Building;
- Fire Zone 1.3, Reactor Building 17 ft, High Pressure Coolant Injection (HPCI)
Pump/Turbine Room;
- Fire Zone 5.6, Electric Fire Pump and Open Areas of Intake, and Yard Areas;
- Fire Zone 3.2, Radwaste & Control Building 23 ft, Cable Spreading Room;
- Fire Zone 1.5, Reactor Building, RCIC Room;
- Fire Zone 1.1, Reactor Building, A RHR and Core Spray Room;
- Fire Zone 1.6, Reactor Building, Control Rod Drive Pump Room;
- Fire Zone 2.1, Radwaste & Control Building 23 ft, B Switchgear and Load Center Room;
- Fire Zone 1.9A, A RHR Valve Room;
- Fire Zone 1.2, B RHR & Core Spray Quadrant;
- Fire Zone 1.4, HPCI Control Panel; and
- Fire Zone 1.7, RCIC Mezzanine.
b. Findings
No findings of significance were identified.
.2 Annual Fire Drill Observation
a. Inspection Scope
(2 samples)
The inspector observed two unannounced fire drills:
- Fire in a cable tray, Reactor Building 23 ft elevation; and
- Fire in the elevator machine room, Operations and Maintenance (O&M) Building.
The fire drills were conducted in accordance with plant procedure ENN-DC-189, Fire Drills. The inspector observed performance of the fire brigade personnel, and confirmed that the licensees fire fighting pre-plan strategies were utilized, the pre-planned drill scenario was followed, and the drill objectives were met. The inspector confirmed that, as appropriate, proper security and radiological controls were applied; proper protective clothing and breathing apparatus were donned; sufficient fire fighting equipment was brought to the scene; the fire brigade leaders fire fighting directions were clear; and communications with the plant operators and between fire brigade members were effective. The inspector confirmed the drill critique identified areas to enhance fire brigade performance. The inspector verified that the licensee identified appropriate corrective actions for identified deficiencies and entered the issues into the corrective action program. This inspection activity represents two samples.
b. Findings
No findings of significance were identified.
1R06 Flood Protection Measures
a. Inspection Scope
(1 sample)
On January 31, 2007, the inspector reviewed protective measures in place to protect against internal flooding of the A and B EDG rooms. The inspector performed visual inspections of the water scuppers on the perimeter of the EDG building to determine whether there were obstructions. Building floor drains were inspected to determine whether there were blockages. Flood barriers which separate the A and B EDG compressor rooms were inspected to ensure they could perform their intended function.
This inspection activity represents one sample for internal flood protection.
b. Findings
No findings of significance were identified
1R11 Licensed Operator Requalification Program
Resident Inspector Quarterly Review
a. Inspection Scope
(1 sample)
The inspector observed a licensed operator simulator exam given on January 23, 2007.
The exam was administered using scenario SES-047, Revision 1, and involved both operational transients and design basis events. The inspector verified that simulator conditions were consistent with the scenario and reflected the actual plant configuration (i.e., simulator fidelity). The inspector observed the crews performance to determine whether the crew met the scenario objectives, accomplished the critical tasks, demonstrated proper use of abnormal and emergency operating procedures, demonstrated proper command and control, and communicated effectively. The inspector observed the evaluators post-scenario critique and confirmed items for improvement were identified and discussed with the operators to further enhance performance. This inspection activity represents one sample.
b. Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness
a. Inspection Scope
(3 samples)
The inspector reviewed the follow-up actions for selected system, structure, or component (SSC) issues and reviewed the performance history of these SSCs to assess the effectiveness of Entergys maintenance activities. The inspector reviewed Entergys corrective actions for these issues in accordance with Pilgrim procedures and the requirements of 10 CFR 50.65(a)(1) and (a)(2), Requirements for Monitoring the Effectiveness of Maintenance. In addition, the inspector reviewed selected SSC classification, performance criteria and goals, system health reports, and corrective actions that were taken or planned to verify whether the actions were reasonable and appropriate. The inspector attended licensee meetings and reviewed licensee plans to address the systems in maintenance rule (a)(1) status. The following issues were reviewed:
- Classification of equipment failures for System 48, the standby gas treatment (SBGT) system. The inspector reviewed licensee actions for condition reports (CRs) pertaining to the SBGT System Instrument Air System, including CRs
===200402377, 200500784, 200501208, 200501250, and 200602645. The inspector reviewed the licensees basis for placing the system in maintenance rule (a)(2) status.
- Classification of equipment failures for System 61, the SBODG. The inspector reviewed licensee actions for select condition reports pertaining to the SBODG system, including CR 20070084. The inspector reviewed the licensees plans for returning the system to (a)(2) status by October 2007.
- Classification of equipment failures for System 61, the station EDG. The inspector reviewed licensee actions for select condition reports pertaining to the EDG system, including CRs 20070056, 20070282, 20070552 and 20070703.
The inspector discussed with licensee staff the plans for returning the system to (a)(2) status by September 2007.
This inspection activity represents three samples.
b. Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control
a. Inspection Scope
=
The inspector evaluated on-line risk management for planned and emergent work. The inspector reviewed maintenance risk evaluations, work schedules, recent corrective actions, and control room logs to verify that concurrent planned and emergent maintenance or surveillance activities did not adversely affect the plant risk already incurred with the out-of-service components. The inspector evaluated whether Entergy took the necessary steps to control work activities, minimize the probability of initiating events, and maintain the functional capability of mitigating systems. The inspector assessed Entergys risk management actions during plant walkdowns. The inspector also discussed the risk management with maintenance, engineering, and operations personnel, as applicable, for the activities. References used for the inspection are identified in the attachment to this report. The inspection activity represents seven samples.
- Planned valve diagnostics on Reactor Building Closed Cooling Water (RBCCW)valve 4060B on January 16, 2007;
- Emergent work on a horizontal support insulator for the offsite 345 kV [kilovolt]
Line 355 on January 18, 2007;
- Emergent work on Reactor Building isolation valve AO-N-91 on February 8, 2007;
- Planned calibration of 345 kV watthour transducers - control room telemetering instruments on January 26, 2007;
- B EDG emergent work on January 4, 2007;
- B EDG governor replacement emergent work February 23, 2007; and
- A EDG emergent work on March 14, 2007.
b. Findings
No findings of significance were identified.
1R15 Operability Evaluations
a. Inspection Scope
(4 samples)
The inspector reviewed selected operability determinations to assess the adequacy of the evaluations, the use and control of compensatory measures, compliance with the TS, and the risk significance of the issues. The inspector used the TS, UFSAR, associated design basis documents, and the additional references listed in the attachment to this report. This inspection activity represents four samples.
- CR 20070108, Reasonable Expectation for Operability (REO) for Drywell Leakage Measurement System;
- CRs 20070293, 20070703, B EDG Load Swings - Operability Evaluation;
- CR 20070629, B18 Motor Control Center and VRC-203A Air Conditioning Unit; and
- CR 20070670, Dual Indication on HPCI Steam Isolation Valve MO-2301-3.
b. Findings
No findings of significance were identified. NRC review of the B EDG is discussed further in Section 1R19 below.
1R17 Permanent Plant Modifications
a. Inspection Scope
(1 sample)
The inspector selected a risk-significant plant modification package for review to verify that the design bases, licensing bases, and performance capability of the risk significant system had not been degraded through the modification. The modification selected for review was Engineering Request 03118068, which restored automatic pumping of the drywell sumps.
For the selected modification, the inspector reviewed the design assumptions and validations to determine the design adequacy. In addition, the inspector reviewed the associated 10 CFR 50.59 safety evaluation to verify that the safety issue pertinent to the changes was properly resolved or adequately addressed. The inspector also reviewed:
- (1) field implementation of the changes to the drywell sump pump controls;
- (2) post-modification functional testing to determine the readiness for operations; and,
- (3) compensatory measures used to monitor reactor coolant system leakage. The inspector reviewed the associated drawings to independently verify the changes and post-work test methods were appropriate. The inspector walked-down portions of the modification on radwaste panel C20 and observed the status of indications on control room panels C903 and C904. The inspector monitored the performance of the drywell leakage system during periodic reviews of plant operations. The inspector reviewed the changes to procedure 2.2.125.1, Reset of Primary and Secondary Containment Isolations (Group I, II, III, IV, V, VI and VII), which were to enhance the administrative control of the sump pumps and isolation valves after a containment isolation.
References used during this review as listed in the attachment to this report. This inspection activity represents one sample.
b. Findings
No findings of significance were identified.
1R19 Post-Maintenance Testing
a. Inspection Scope
(5 samples)
The inspector reviewed post-maintenance test (PMT) activities on risk significant systems to determine whether the effect of the test on the plant had been evaluated adequately, the test was performed in accordance with procedures, the test data met the required acceptance criteria, and the test activity was adequate to verify system operability and functional capability following maintenance. The inspector confirmed that systems were properly restored following testing and that discrepancies were appropriately documented in the corrective action process. References used during this review are listed in the attachment to this report. The inspection activity represents five samples.
- MR 07102225, Post Work Test for Refuel Floor Isolation Valve AO-N-91;
b. Findings
Introduction.
A Green self-revealing NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified when Entergy failed to promptly correct a condition adverse to quality associated with the B EDG. As a result, the EDG was not capable of performing its safety function for 29 days.
Description.
On January 21, 2007, the B EDG was taken out of service for a planned maintenance overhaul. The overhaul consisted of inspections and preventive maintenance of the B EDG and support systems. One of the maintenance tasks was a flush and change of the B EDG governor oil. On January 25, 2007, following the completion of the overhaul, Entergy began a B EDG operability test which included a full load run to 2600 kW, in accordance with Entergy procedure 8.9.1, Emergency Diesel Generator and Associated Emergency Bus Surveillance. During the operability run, operators noted unexpected oscillations in both kilowatts and generator amperage.
Kilowatt oscillations measured approximately 150 kW and generator amperage oscillations measured approximately 10 to 12 amperes. Procedure 8.9.1 states that if kilowatt oscillations are greater than 200 kW, then shutdown of the diesel engine is required. Although not exceeding the 200 kW abort criteria, Entergy entered the degraded condition in their corrective action program as CR 20070293. Entergy concluded that the oscillations were minor and, since they did not exceed the abort criteria in procedure 8.9.1, the operability run was successful and the B EDG was operable. Corrective actions planned for the kW oscillations included instrumenting the B EDG with a test recorder to obtain data to tune the governor in order to achieve a more stable operation of the B EDG. The corrective actions were scheduled to coincide with the next surveillance run on February 23, 2007, which was 29 days after the initial kW oscillations were observed.
During the February 23, 2007, B EDG operability run, Entergy again noted kW oscillations at a full load of 2600 kW. Entergy noted this condition in their corrective action program in CR 20070703. Specifically, approximately 50 minutes into the full load portion of the operability run, kW oscillations were observed at 100 kW and continued to increase during the run to 200 kW at a full load of 2600 KW. The full load run portion of the test was completed and the control room operator started to unload the engine in accordance with the procedure; as the load was reduced to 1800 kW, the kW oscillations were observed to exceed 200 kW. At this time, the control room supervisor ordered that the B EDG be secured due to the procedure abort criteria of greater than 200 kW being exceeded. Entergys subsequent investigation included reviewing the test recorder data and sampling the oil from the mechanical portion of the governor (referred to as the EGB). Entergy found that the sampled oil from the EGB was gray and cloudy, and had a burnt smell, unlike the oil that was installed in the governor during the January 2007 overhaul. Initial testing of the EGB oil by the site Chemistry department indicated the presence of unidentified contaminants. Based largely on the condition of the governor oil sample, Entergy concluded that the EGB required replacement.
Entergy Maintenance replaced the EGB on February 24, 2007, and tuned the governor system in accordance with Entergy procedure 3.M.3-61.7, Woodward Governor Tuning. The governor oil system was vented, and on February 25, 2007, an operability run was commenced. During the operability run, load oscillations did not exceed 30 kW.
Entergy further tested the governor by performing a load-reject test of the B EDG. The purpose of this test was to monitor the response of the governor control system when a large load (core spray pump) was started and tripped. This test was satisfactorily completed. The B EDG was then declared operable.
On March 7, 2007, Entergy shipped the EGB to the vendor for testing and failure analysis. When tested by the vendor, the EGB lost pressure within a few minutes and the test was aborted. The vendor performed an inspection of the EGB and found shreds of aluminum in the governor oil. The source of the aluminum was determined to be an aluminum label that was originally attached to the shutdown solenoid valve located inside the EGB. Entergy concluded that the most likely scenario was that the label was installed during maintenance in 2002 and became dislodged as a result of the governor oil change during the January 2007 engine overhaul. The label was then ground up by the relay bushing gears. The ground particles then clogged the internal passages of the governor, resulting in a degraded governing ability of the EGB. This resulted in the unexpected kW oscillations of the B EDG.
Entergy conducted an extent of condition review and tested the oil from the A EDG governor. Test results showed that the oil in the A EDG governor was acceptable with no aluminum particles present. The root cause investigation concluded that this issue was isolated to the B EDG. Entergy analyzed the effects that the 200 kW oscillations would have had on connected emergency core cooling system loads while the B EDG operated. The frequency changes resulting from 200 kW oscillations were found not to cause any premature tripping of loads or other adverse consequences.
Analysis.
The inspector determined that the failure to promptly correct a condition adverse to quality associated with the B EDG was a performance deficiency. While the kW swings during the January test did not exceed the 8.9.1 abort criteria, the oscillations were not typical of historical performance. The inspector concluded that the B EDG was inoperable for 29 days because the EGB function was impacted with an ongoing degradation mechanism. The inspector determined that this finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the reliability, availability, and capability of systems that respond to initiating events to prevent undesirable consequences.
The finding was determined to be of very low safety significance (Green) in accordance with Inspection Manual Chapter (IMC) 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspector initially evaluated the significance of this finding using IMC 0609, Appendix A, and determined that a Phase 2 analysis was required since the finding represented an actual loss of safety function of a single train for greater than the TS allowed outage time. An NRC Senior Reactor Analyst (SRA) determined a Phase 3 analysis was necessary due to the indication of a greater-than-green finding using the site-specific Phase 2 notebook. The SRA used the Pilgrim Standardized Plant Analysis Risk (SPAR) model to analyze the finding and noted differences in how the Phase 2 notebook and the SPAR model credited the use of the 23kV offsite power source and the SBODG. The SPAR model dominant cutsets are a station blackout with the HPCI train unavailable and a stuck open safety relief valve. The SRA determined that the SPAR model correctly characterized the significance of the finding. The SRA further evaluated the potential risk contribution from large early release frequency and the contribution from external events including seismic and fire initiated loss of offsite power events. Based on this review, the SRA concluded that the finding was of very low safety significance (Green).
The inspector determined that this finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, in that Entergy personnel failed to thoroughly evaluate the unexpected kW oscillations.
Enforcement.
10 CFR 50, Appendix B, Criterion XVI, Corrective Action, requires that measures be established to assure that conditions adverse to quality are promptly identified and corrected. Contrary to the above, on January 25, 2007, Entergy failed to promptly correct a condition adverse to quality when load swings of 150 kW were found during a post maintenance operability run of the B EDG. Because this violation is of very low safety significance and has been entered in Entergys corrective action program (CR 20070703), this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000293/20070201, Inadequate Evaluation of Unexpected Emergency Diesel Generator Load Swings.
1R20 Refueling and Other Outage Activities
a. Inspection Scope
(2 samples)
The inspector reviewed Entergys plans and preparations for the refueling outage scheduled to begin in April 2007. This inspection activity represents two samples.
- Review of Refuel Outage Plan. The inspector reviewed the refueling outage (RFO) -16 Outage Shutdown Risk Assessment and procedure TP07-023, Compensatory Measures, to verify that Entergy addressed the outage impact on defense-in-depth for the five shutdown critical safety functions: electrical power availability, inventory control, decay heat removal, reactivity control, and containment. The inspector reviewed how Entergy planned to provide adequate defense-in-depth for each safety function, and the planned contingencies to minimize the overall risk where redundancy was limited or not available.
Consideration of operational experience was also assessed.
- New Fuel Receipt and Inspections. The inspector observed licensee activities to receive and inspect new fuel for Operating Cycle 17, install fuel channels, and store the fuel in the spent fuel pool. The inspector used the following references for the review: Procedure 4.1, Receiving and Handling of Unirradiated Fuel Assemblies; Procedure 4.2, Inspection and Channeling of Nuclear Fuel; TS 3.7, Containment Systems; and UFSAR Section 10.3, Spent Fuel Storage.
b. Findings
No findings of significance were identified.
1R22 Surveillance Testing
a. Inspection Scope
(8 samples)
The inspector observed surveillance tests and/or reviewed test results to determine whether the test acceptance criteria were consistent with TS, that the tests were performed in accordance with the written procedure, that the test data was complete and met procedural requirements, and the components were capable of performing their intended safety functions. Additional references used for this review are listed in the attachment to this report. The inspection activity represents eight samples.
- 8.5.5.9, RCIC Simulated Automatic Actuation, Flow Rate, and Cold Quickstart Test;
- 8.9.1, A EDG Monthly Surveillance;
- 8.M.2-1.5.8.3, Logic System Functional Test of System A Standby Gas Treatment Initiation, Reactor Building Isolation and Inboard Drywell Isolation Valves (Atmospheric Control Valves);
- 7.3.36, Offgas Sampling and Air Ejector Radiation Monitor Setpoints;
- 8.5.3.1, RBCCW System Quarterly and Biennial Comprehensive Operability (P-202D);
- 8.5.1.1, Core Spray Pump Biennial IST [Inservice Test]; and
- 8.A.18, Core Spray System Integrity Surveillance.
b. Findings
No Findings of significance were identified. Licensee identified findings are described in Section 4OA7 of this report.
1R23 Temporary Plant Modifications
a. Inspection Scope
(2 samples)
The inspector reviewed the temporary modifications identified below to verify that the licensing bases and performance capability of the associated risk significant system had not been degraded through the modification. A walkdown was performed to determine whether equipment was installed in accordance with instructions. The inspector reviewed applicable drawings and procedures to determine whether they properly reflected the temporary modifications. The references used for this review are listed in the attachment to this report. This inspection activity represents two samples.
- Temporary Alteration 06-1-064, Drill Through Disc of 29-HO-3A; and
- Temporary Alteration 06-1-063, Plexiglass in Screen House.
b. Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP6 Drill Evaluation
a. Inspection Scope
(1 sample)
The inspector observed an evaluated licensed operator simulator training exercise on January 23, 2007, and evaluated the crews ability to implement the emergency plan.
Specifically, the inspector confirmed that the crew properly classified the event, activated the notification system, and appropriately completed and transmitted the event notification forms in a timely manner. This inspection activity represents one sample.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
[OA]
4OA1 Performance Indicator Verification
.1 Reactor Safety Cornerstones
a. Inspection Scope
(3 samples)
The inspector reviewed Performance Indicator data to determine the accuracy and completeness of the reported data. The review was accomplished by comparing reported Performance Indicator data to confirmatory plant records and data available in plant logs, the chemistry data base, maintenance rule records, Licensee Event Reports, condition reports, and NRC inspection reports. The inspection activity represents three samples.
- Barrier Integrity Cornerstone, Reactor Coolant System Unidentified Leakage from the second quarter of 2006 through the first quarter 2007;
- Barrier Integrity Cornerstone, Reactor Coolant System Specific Activity from the second quarter of 2006 through the first quarter 2007; and
- Mitigating System Cornerstone, Safety System Functional Failures from the first quarter of 2006 through the first quarter 2007.
b. Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems
Reactor Safety Cornerstone
.1 Review of Items Entered into the Corrective Action Program
a. Inspection Scope
As required by Inspection Procedure 71152, Identification and Resolution of Problems, the inspector performed a screening of each item entered into the licensees corrective action program. This review was accomplished by reviewing printouts of each condition report, attending daily screening meetings, and/or accessing the licensees database.
The purpose of this review was to identify conditions, such as repetitive equipment failures or human performance issues, that might warrant additional follow-up.
b. Findings
No findings of significance were identified.
.2 Annual Sample - Station Blackout Diesel Generator
a. Inspection Scope
(1 sample)
The inspector reviewed Entergys corrective actions for the failure of the SBODG to start on demand in January 2005. Specifically, on January 23, 2005, following the loss of electrical power to its auxiliary equipment, operations attempted to start the SBODG, as specified in procedure 2.4.146, Station Blackout Diesel Generator, when jacket water outlet temperature dropped to 80EF. The diesel generator failed to start, tripping on low lube oil pressure. The inspector reviewed responses to CRs 200501177, 200500392, and 200500256; and the procedure changes specified for procedures 2.2.146, Station Blackout Diesel Generator; 2.4.16, Distribution Alignment Electrical System Malfunctions; and 2.4.A.23, Loss/Degradation of 23 kV Line.
b. Findings and Observations
No findings of significance were identified. However, a negative observation regarding Engineering Department performance was noted during this review. Engineerings original proposed solution, to increase the jacket water outlet temperature from 80EF to 85EF, lacked an adequate basis to assure the diesel engine would start. After prompting by Operations Department personnel, the issue was re-evaluated and Entergy determined that the appropriate corrective action was the revision of station procedures to require starting the SBODG as soon as possible following a loss of power to its auxiliary equipment. Prior testing had demonstrated the SBODG was able to start 10 minutes following loss of power to its auxiliary equipment. This inspection activity represents one sample.
4OA3 Event Follow-up
a. Inspection Scope
(4 samples)
The inspector assessed the control room operator performance during the following planned and un-planned non-routine evolutions. The inspector evaluated personnel performance based on control room observations, interviews, and reviews of operator logs, alarm response procedures, and operating procedures. This inspection activity represents four samples.
- The inspector observed a planned plant power reduction to 62 percent power on January 26-27, 2007, per procedure 2.1.14, Station Power Changes, and power suppression testing, per procedure 9.32, Power Suppression Testing.
The inspector used power maneuvering plan MAN.C16-76 as a reference for this review. The inspector reviewed Entergys actions to identify a potential fuel leak, suppress the fuel defect, and follow the fuel pre-conditioning operating guidelines.
- The inspector reviewed the operator response to a partial loss of the 23 kV power supply on January 26, 2007, due to a line failure on the site access road.
The inspector reviewed the operator actions per procedure 2.4.A.23, Loss/Degradation of 23 kV Line, and the actions to restore plant conditions following the restoration of power.
- The inspector responded to the site on March 17, 2007, to review Entergys actions to shutdown and cooldown the plant in response to increasing drywell leakage. During plant operations at 86 percent power in end-of-cycle coast down on March 17, 2007, the operators noted unidentified drywell leakage increased from 0.57 gallons per minute (gpm) to 1.16 gpm over a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> period. The operators began a controlled shutdown at 12:30 p.m. to place the plant in cold shutdown in order to investigate and repair the leakage source. The operators reduced reactor power using recirculation flow and control rods, and manually scrammed the reactor from 30 percent power at 4:55 p.m. on March 17, 2007. The reactor responded as expected during the scram and safety systems operated properly. Reactor Building and containment isolations and RWCU system isolation occurred as expected due to low reactor vessel level following the scram [Event Number 43235]. Drywell unidentified leak rate increased to a maximum value of 2.59 gpm, and then decreased as the reactor was cooled down and depressurized. Drywell leakage remained below the TS 3.6.C.1 limits for total (25 gpm) and unidentified (5 gpm) leak rate.
The plant entered cold shutdown at 4:32 pm on March 18, 2007. Entergy personnel entered the drywell and identified a packing leak from RWCU system valve MO-1201-85, which was subsequently repaired (MR 07104403). This item was entered into Entergys corrective action program as CR 20070949 to investigate the root cause of the packing leak.
During the above sequence of events, the inspector observed the operators reduce power per power maneuvering plan MAN.C16-84 and shutdown the reactor by inserting a manual scram per procedure 2.1.6, Reactor Scram. The inspector observed Entergy actions to implement plant cooldown activities using procedure 2.1.5, Controlled Shutdown from Power, and 2.1.7, Vessel Heatup and Cooldown. The inspector reviewed Entergys scram report and verified that the cause of the shutdown was understood and corrected prior to plant restart.
Additional references used during this review are listed in the attachment to this report.
- The inspector reviewed Entergy actions on March 18-19, 2007, to restart the plant using procedures 2.1.1, Startup from Shutdown, and 2.2.70, Drywell and Torus Inerting, and power maneuvering plan MAN.C16-84. The startup activities observed included the approach to criticality, rolling the main turbine, and synchronizing the unit to the electrical grid. The reactor was restarted at 3:08 p.m. on March 19, 2007. The unidentified leak rate in the drywell was 0.03 gpm following the resumption of plant power operations. The inspector reviewed Entergy actions to enter deficiencies into the corrective action program. The references used during this review are listed in the attachment to this report.
b. Findings
No findings of significance were identified.
4OA6 Meetings, Including Exit
Exit Meeting Summary
On April 5, 2007, the inspection results were presented to Mr. Kevin Bronson, and members of his staff. The inspector asked the licensee whether any of the material examined during the inspection should be considered proprietary. No proprietary information was identified.
4OA7 Licensee-Identified Violations
The following violations of very low safety significance (Green) were identified by Entergy and are violations of NRC requirements which meet the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a Non-Cited Violation.
- TS 5.4.1 requires, in part, that Entergy establish and implement procedures to cover the activities specified in Regulatory Guide 1.33, Appendix A. Contrary to the above, the requirements contained in procedures 8.5.3.1, Reactor Building Closed Cooling Water System Quarterly and Biennial Comprehensive Operability, and 1.3.34.2, Limiting Conditions For Operation Log, were not implemented on February 24, 2007, for a failed surveillance test of the D RBCCW pump (P-202D). Specifically, operators did not identify the pumps total dynamic head of 81.5 ft was below the minimum allowed total dynamic head of 82 ft. The error was identified by system engineering on February 26, during review of the completed procedure. Condition Reports 200700719 and 200700737 were issued and operations entered P-202D into tracking LCO (Track-1-07-0016), as required by procedure 1.3.34.2. The finding is more than minor per MC-0612, Appendix E, examples 2a and 2c, in that the total dynamic head was below the minimum specified value. The issue is not greater than Green (very low safety significance) because the RBCCW loop remained operable due to the availability of the two remaining pumps.
- TS 5.4.1 requires, in part, that procedures be established and implemented covering the activities specified in Regulatory Guide 1.33, Appendix A. Contrary to the above, plant personnel did not implement the requirements of procedures 7.3.36, Offgas Sampling and Analysis, and 7.4.63, Process Radiation Monitoring Setpoints, when samples of the offgas system indicated the setpoints for offgas radiation monitors 1705-3A/B were non-conservative. The setpoints are used to assure the controls of Offsite Dose Calculation Manual (ODCM) 3.1.2 are met. The discrepancy was identified after 26 samples had been taken from January 16, 2007, through March 3, 2007. The licensee entered action statement LCO-ACT-1-07-035 and completed actions to adjust the setpoints on March 3, 2007. The issue was more than minor per MC-0612, Appendix A, because the non-conservative setpoints impact the public radiation safety cornerstone which involves radiological effluent monitoring and was contrary to the controls in the ODCM 3.1.2. The issue was not more than Green because offgas release rates remained well below the levels specified in TS 3.8.1, and other radiation monitors (post-treat and main stack) in the effluent pathway were operable and would have alerted the operators of the need for action. This discrepancy was entered into the Entergys corrective action system as CR 20070784.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
- K. Bronson Site Vice President, Pilgrim
B. Cobb Maintenance Supervisor
T. Collis System Engineer
W. Cook Supervisor Electrical Engineering
F. Dicristofara Senior Operations Specialist
D. Ellis Senior Engineer
J. Fitzsimmons Radiation Protection Supervisor
B. Ford Licensing Manager
G. James Reactor Engineering Superintendent
J. Keenan System Engineer
T. McElhinney Chemistry Superintendent
C. McMorrow Senior Operations Instructor
J. Moylan Electrical Supervisor
D. Noyes Assistant Operations Manager
E. Olson Operations Manager
M. Santiago Training Superintendent
K. Sejkora Senior Chemistry Specialist
R. Smith General Manager-Plant Operations
NRC personnel
- W. Raymond, Senior Resident Inspector
- C. Welch, Resident Inspector
LIST OF ITEMS
OPENED, CLOSED AND DISCUSSED
Open and
Closed
- 05000293/20070201 NCV Failure to Thoroughly Evaluate Degraded Condition on B EDG Following January Overhaul (Section 1R19)