IR 05000333/2011002
| ML111240331 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 05/04/2011 |
| From: | Mel Gray Reactor Projects Branch 2 |
| To: | Bronson K Entergy Nuclear Northeast |
| Gray, Mel NRC/RGNI/DRP/PB2/610-337-5209 | |
| References | |
| IR-11-002 | |
| Download: ML111240331 (24) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION I
475 ALLENDALE ROAD
SUBJECT:
JAMES A. FITZPATRICK NUCLEAR POWER PLANT - NRC INTEGRATED I NSPECTION REPORT 05000333/201 1002
Dear Mr. Bronson:
On March 31,2011, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your James A. FitzPatrick Nuclear Power Plant (FitzPatrick). The enclosed inspection report dotuments the inspection results which were discussed on Aprll27 ,2011, 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 Commission's 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 resutts of this inspection, this report documents one NRC-identified finding of very low safety significance (Green). This finding was determined to involve a violation of NRC requiremlntJ. Howevei, because of the very low safety significance and because the issue is eniered into your corrective action program, the NRC is treating this finding as a non-cited violation (NCV) consistent with Section 2.3.2 of the NRC's Enforcement Policy. lf you contest this NCV, you should provide a response within 30 days of the date of the inspection report, with the basis for your denial, to the U. S. Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington, D.C. 20555-0001; with a copy to the Regional Administrator, Region l: Office of Enforcement; U. S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Senior Resident Inspector at FitzPatrick. In addition, if you disagree with the cross-cutting aspect assigned to the 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 RegionalAdministrator, Region l, and the NRC Senior Resident Inspector at FitzPatrick.
K. Bronson 2 ln accordance with 10 CFR Part 2.390 of the NRC's "Rules of Practice," a copy of this letter and 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 NRC's document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/readinq-rmiadams.html (the Public Electronic Reading Room).
Sincerely, rtl4"M Mel Gray, Chief Projects Branch 2 Division of Reactor Projects Docket No.: 50-333 License No.: DPR-59 Enclosure: I nspection Report 05000333/201 1 OO2 w/Attachment: Supplemental I nformation cc w/encl: Distribution via ListServ
SUMMARY OF FINDINGS
lR 0500033312011002:0110112011 - 0313112011; James A. FitzPatrick Nuclear Power Plant; ldentification and Resolution of Problems.
The report covered a three-month period of inspection by resident inspectors and an announced inspection by a region-based inspector. One Green finding, which was a non-cited violation, was identified. The significance of most findings is indicated by their color (Green, White,
Yellow, Red) using Inspection Manual Chapter (lMC) 0609, "Significance Determination Process" (SDP). The cross-cutting aspect for the finding was determined using IMC 0310,
"Components Within the Cross-Cutting Areas." Findings for which the SDP does not apply may be "Green" or be assigned a severity level after Nuclear Regulatory Commission (NRC)management review. The NRC's 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 a non-cited violation (NCV) of very low safety signriticance of 10 CFR 50, Criterion XVl, "Corrective Action," because Entergy personnel did not identify and correct a condition adverse to quality related to a control room envelope (CRE) boundary door. Specifically, Entergy personnel did not identify and implement adequate actions to ensure the safety-related CRE boundary door, 70DOR-4-300-5, remained latched and able to perform its safety function. As corrective action, the foreign material that prevented the door from consistently latching was removed by Entergy personnel. The issue was entered into the corrective action program (CAP) as condition reports CR-JAF-2010-08617 and CR-JAF-2011-00407'
The finding was more than minor because it was associated with the configuration control and the barrier performance attributes specific to the radiological barrier function of the control room. The finding affected the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. The finding was determined to be of very low safety significance in accordance with tMC 0609, Appendix A, "Determining the Significance of Reactor Inspection Findings for At-Power Situations," based on a Phase 3 analysis. The inspectors determined the period that the door was potentially open was small relative to the technical specification (TS) allowed outage time, and therefore represented very low safety significance, considering the low probability of a design basis accident during that time period.
The finding had a cross-cutting aspect in the area of problem identification and resolution within the corrective action program component because Entergy personnel did not completely and accurately identify the degraded condition of the door (P.1(a) per IMC 0310). (Section 4OA2)
REPORT DETAILS
Summarv of Plant Status The James A. FitzPatrick Nuclear Power Plant (FitzPatrick) began the inspection period operating at 100 percent reactor power. On March 21 , operators reduced reactor power to 55 percent ior a control rod sequence exchange, single control rod scram time testing, to identify and plug main condenser tube leaks, and to repair a leak from the 'A' reactor feedwater pump.
Operators restored reactor power to 100 percent on March 24. The plant continued to operate at or near full power for the remainder of the inspection period'
1. REACTOR SAFEW
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R04 Equipment Aliqnment
.1 Quarterlv Partial Svstem Walkdown
a.
Inspection ScoPe The inspectors performed four partial system walkdowns to verify the operability of redundant or diverse trains and components during periods of system train unavailability or following periods of maintenance. The inspectors referenced system procedures, the updated final safety analysis report (UFSAR), and system drawings in order to verify the aiignment of the available train was proper to support its required safety functions. The inspectors also reviewed applicable condition reports (CRs) and work orders (WOs)to ensure that Entergy personnel identified and properly addressed equipment discrepancies that could impair the capability of the available equipment train, as required by Title 10, Code of Federal Regulations (10 CFR) Part _50, Appendix B, Criierion XVl, "Corrective Action." The documents reviewed are listed in the Attachment.
The inspectors performed a partial walkdown of the following systems:
.
'A' residual heat removal service water (RHRSW) when 'B' RHRSW was out of service for maintenance; o Reactor core isolation cooling (RCIC) when high pressure coolant injection (HPCI)was unavailable due to surveillance testing;
.
'A' standby liquid control (SLC) while'B' SLC was out of service for maintenance; and
.
'B'emergency service water (ESW) while 'A' ESW was out of service for maintenance.
These activities constituted four partial system walkdown inspection samples.
b. Findinqs No findings were identified.
.2 Complete Svstem Walkdown
a. Inspection Scope
The inspectors performed a complete system alignment inspection of the HPCI system to identify discrepancies between the existing equipment lineup and the required lineup.
During the inspection, system drawings and operating procedures were used to verify proper equipment alignment and operational status. The inspectors reviewed the open maintenance WOs associated with the system for deficiencies that could affect the ability of the system to perform its function. Documentation associated with unresolved design issues such as temporary modifications, operator workarounds, and items tracked by plant engineering were also reviewed by the inspectors to assess their collective impact on system operation. In addition, the inspectors reviewed the corrective action program (CAP) database to verify that equipment problems were being identified and appropriately resolved. The documents reviewed are listed in the Attachment.
These activities constituted one complete system walkdown inspection sample.
b. Findinqs No findings were identified.
1R05 Fire Protection
.1 Quarterlv Review
a. Inspection Scope
The inspectors conducted inspections of fire areas to assess the material condition and operational status of fire protection features. The inspectors verified, consistent with applicable administrative procedures, that combustibles and ignition sources were adequately controlled; passive fire barriers, manualfire-fighting equipment, and suppression and detection equipment were appropriately maintained; and compensatory measures for out-of-service, degraded, or inoperable fire protection equipment were implemented in accordance with FitzPatrick's fire protection program. The inspectors evaluated the fire protection program for conformance with the requirements of license condition 2.C(3), "Fire Protection." The documents reviewed are listed in the
.
r Turbine building miscellaneous oil storage room, fire arealzone lE/OR-3;
.
Reactor building (RB) 272 foot elevation, fire arealzone lX/RB-1A, XRB-18;
.
RB 300 foot elevation, fire arealzone Vlll/RB-1C, IXRB-1A, )URB1B; r East cable tunnel, fire arealzone ll/CT-2; and
.
Crescent area - west, fire arealzone XVlll/RB-1W' These activities constituted five quarterly fire protection inspection samples.
b. Findings
No findings were identified.
1R06 Flood Protection Measures
a. Inspection Scope
The inspectors conducted tours of the relay room to assess internal flooding protection measures in that area. The inspectors reviewed selected risk significant plant design features intended to protect the associated safety-related equipment from internal flooding events. The inspectors reviewed flood analysis and design documents, including the lndividual Plant Examination and UFSAR.
These activities constituted one internalflood protection measures inspection sample.
b. Findinqs No findings were identified.
1R11 Licensed Operator Requalification Proqram
a. Inspection Scope
On March 7, the inspectors observed licensed operator simulator training to assess operator performance during a scenario to verify that crew performance was adequate and evaluators were identifying and documenting crew performance problems. The inspectors evaluated the performance of risk significant operator actions, including the use of emergency operating procedures. The inspectors assessed the clarity and effectiveness of communications, the implementation of appropriate actions in response to alarms, the performance of timely control board operation and manipulation, and the oversight and direction provided by the shift manager. Licensed operator training was evalualed for conformance with the requirements of 10 CFR Part 55, "Operators' Licenses." The documents reviewed are listed in the Attachment.
These activities constituted one quarterly operator simulator training inspection sample.
b.
Findinqs No findings were identified.
1R12 Maintenance Effectiveness
a.
Inspection ScoPe The inspectors reviewed performance-based problems involving selected in-scope structures, systems, or components (SSCs) to assess the effectiveness of the maintenance program. The reviews focused on the following aspects when applicable:
o Proper maintenance rule scoping in accordance with 10 CFR 50.65;
.
Characterization of reliability issues;
.
Changing system and component unavailability;
.
10 CFR 50.65 (aX1) and (a)(2) classifications;
.
ldentifying and addressing common cause failures;
.
Appropriateness of performance criteria for SSCs classified (aX2); anO
.
Adequacy of goals and corrective actions for SSCs classified (aX1).
The inspectors reviewed system health reports, maintenance backlogs, and Maintenance Rule basis documents. The documents reviewed are listed in the
. The following systems were selected for review.
r High pressure coolant injection system; and o Automatic depressurization system.
These activities constituted two quarterly maintenance effectiveness inspection samples.
b. Findinqs No findings were identified.
1R13 Maintenance RiskAssessments and EmerqentWork Control
a. lnspection Scope The inspectors reviewed maintenance activities to verify that the appropriate risk assessments were performed prior to removing equipment for work. The inspectors reviewed whether risk assessments were performed as required by 10 CFR 50'65(a)(4),and were accurate and complete. When emergent work was performed, the inspectors reviewed whether plant risk was promptly reassessed and managed. The documents reviewed are listed in the Attachment. The reviews focused on the following activities:
o The week of January 3, that included turbine valve testing, RCIC system quarterly surveillance testing, chemical cleaning of safety related unit coolers in the west crescent room, and emergent maintenance to troubleshoot failure of the RCIC steam admission valve, 13MOV-131, to fully open on demand'
.
The week of January 31, that included 'A' and 'C' emergency diesel generator (EDG)monthly surveillance testing, 'A' residual heat removal (RHR) system quarterly surveillance testing, and emergent maintenance to correct a malfunctioning computer position indication for the 'A' to 'C' EDG cross-tie breaker, Tl-1Q504.
o The week of February 14, that included 'B'and 'D' EDG monthly surveillance testing,
'B' standby gas treatment system maintenance, 'B' station battery ventilation subsystem maintenance, 'B' RHR and RHRSW systems quarterly surveillance testing, and emergent maintenance to troubleshoot an unexpected two percent reduction in 'B' average power range monitor (APRM) indication.
The week of February 28, that included 'A' and 'C' EDG monthly surveillance testing, adjustment of the reactor power high limiter for the 'A' reactor water recirculation motor-generator, 'A' SLC system quarterly surveillance testing, 'A' core spray system quarterly surveillance testing, and maintenance on the'A' battery room ventilation system.
The week of March 21, that included single control rod scram time testing, repair of a leak from the 'A' reactor feedwater pump, torus-to-drywell vacuum breaker qua(erly surveillance testing, and troubleshooting to isolate a main generator field electrical ground.
These activities constituted five maintenance risk assessments and emergent work control inspection samples.
b. Findinqs No findings were identified.
1R15 Operabilitv Evaluations
a. Inspection Scope
The inspectors reviewed operability determinations to assess the acceptability of the evaluations; the use and control of applicable compensatory measures; and compliance with technical specifications (TSs). The inspectors' reviews included verification that the operability determinations were conducted as specified by EN-OP-104, "Operability Determination Process." The technical adequacy of the determinations was reviewed and compared to the TSs, UFSAR, and associated design basis documents (DBDs).
The inspection focused on the following operability reviews:
r CR-JAF-2011-00747 concerning a temperature indicating switch, 92TlS-101A, associated with the 'A' EDG that was maintaining the room temperature at 90 degrees Fahrenheit; e CR-JAF-2011-O}43zconcerning the'C'main steam line radiation monitor, 17RM-251C, with regard to TS-required channel check surveillance and irregular (spiking)indication;
.
CR-JAF-2011-00923 concerning the'B'APRM subsequent to a two percent drop in indicated level that had occurred in association with a power supply replacement for
.A'APRM; r CR-JAF-2O11-00968 concerning the potential effect of a loss of temperature monitoring capability for control rod drive 46-19 on the insertion time for that control rod; and
.
CR-JAF -2011-01230 concerning the effect of the 'A' reactor water recirculation motor-generator high speed mechanical stop position with regard to core flow limitations consistent with core operating limits report (COLR) assumptions with respect to core thermal limits.
These activities constituted five operability evaluation inspection samples.
I b. Findinqs No findings were identified.
1R18 Plant Modifications
a. Inspection Scope
The inspectors assessed the adequacy of the 10 CFR 50.59 evaluations for the following temporary and permanent modifications respectively. The inspectors' reviews considered whether the installations were consistent with the modification documentation, that the drawings and procedures were updated as applicable, and that the post-installation testing was adequate. The following reviews represented one temporary modification inspection sample and one permanent modification inspection sample:
r Temporary modification of the 'A' battery room ventilation system for operation with supply tan 72AHLJ-30A inoperable, in accordance with OP-59A, "Battery Room Ventilation;" and
.
EC 14122, "Modification to lmprove SRV [safety/relief valve] Reliability - Replace 02RV-71 C, -71E, and -71F with Target Rock Three-Stage Safety/Relief Valves."
b. Findinqs No findings were identified.
1R19 Post-Maintenance Testinq
a.
lnspection Scope The inspectors reviewed post-maintenance test procedures and associated testing activities for selected risk-significant mitigating systems to assess whether the effect of maintenance on plant systems was adequately addressed by control room and engineering personnel. The inspectors verified whether test acceptance criteria were clelr, demonstrated operational readiness, and were consistent with DBDs; test instrumentation had current calibrations, adequate range, and accuracy for the application; and tests were performed, as written, with applicable prerequisites satisfied' Upon completion, the inspectors verified whether equipment was returned to the proper alignment necessary to perform its safety function. Post-maintenance testing (PMT) was eviluated for conformance with the requirements of 10 CFR 50, Appendix B, Criterion Xl, "Test Control." The documents reviewed are listed in the Attachment. PMT activities associated with the following work orders were reviewed:
.
WO 00228000, preventive maintenance on the'A' standby gas treatment system; o WO 00260774, replacement of 52STA, stationary auxiliary switch, within the 71-
===10504 cubicle, associated with the tie breaker of the 'A' and 'c' EDGs;
.
WO 52189872, replacement of relay 03A-K50 within the 09-28 panel, associated with rod sequence control logic; o WO 52189878, replacement of relay 03A-K60 within the 09-28 panel, associated with a group notch control insert block;
.
WO 52192021-01, replacement of reactor protection system 81 logic relay 05A-K1048; and r WO 00271482-01, replacement a section of piping in the west crescent service water supply to 'E' and 'G' unit coolers due to a pinhole leak' These activities constituted six PMT inspection samples.
b. Findinos No findings were identified.
1R22 Surveillance Testinq
a. Inspection Scope
The inspectors witnessed performance of surveillance tests (STs) and/or reviewed test data of selected risk-significant SSCs to assess whether the SSCs satisfied TSs, UFSAR, technical requirements manual (TRM), and station procedure requirements.
The inspectors reviewed whether test acceptance criteria were clear, demonstrated operational readiness, and were consistent with DBDs; test instrumentation had current calibrations, adequate range, and accuracy for the application; and tests were performed, as written, with applicable prerequisites satisfied. Upon ST completion, the inspectors verified that equipment was returned to the status specified to perform its safety function. The following STs were reviewed:
.
ST-988, 'EDG B and D Full Load Test and ESW Pump Operability Test," Revision 11; o ST-9AA, "EDG System A Fuel/Lube Oil Monthly Test," Revision 2;
.
ST-9A8, "EDG System B Fuel Oil Monthly Test," Revision 2; r ST-4N, "HPCI Quick-Start, lnservice, and Transient Monitoring Test (lST) [inservice testl," Revision 59; o ST-2AM, "RHR Loop B Quarterly operability Test (lsT)," Revision 29; and o ST-3P8, "Core Spray Loop B Quarterly Operability Test (lST)," Revision 19.
These activities represented six surveillance testing inspection samples' b.
Findinqs No findings were identified.
OTHER ACTIVITIES
4OA2 ldentification and Resolution of Problems (71152 - 1 sample)
.1 Review of ltems Entered into the Corrective Action Proqram
a.
Inspection ScoPe As required by Inspection ProcedureTll52, "ldentifiCation and Resolution of Problems,"
to identify rep-etitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of all items entered into Entergy's CAP' The review was accomplished by accessing Entergy's computerized database for CRs and attending CR screening meetings. In accordance with the baseline inspection procedures, the inspectors selected items across the Initiating Events, Mitigating Systems, Barrier Integrity, and Public Radiation Safety cornerstones for additional fottow-up and review. The inspectors assessed Entergy personnel's threshold for problem identification, the adequacy of the cause analyses, and extent of condition review, operability determinations, and the timeliness of the specified corrective actions.
The CRs reviewed are listed in the Attachment.
b. Findinqs and Observations lntroduction: The inspectors identified an NCV of very low safety significance of 10 CFR 50, Cnterion XVl, "Corrective Action," because Entergy personnel did not identify and correct a condition adverse to quality related to a control room envelope (CRE) boundary door. Specifically, Entergy personnel did not identify and implement adequate actions to ensure the safety-related CRE boundary door, 70DOR-A-300-5, remained latched and able to perform its safety function.
Description:
During a system walkdown performed by Entergy pejs_onnel o.n December 29;;019.2 systernengineer identified door 70DOR-A-300-5, a CRE boundary door between the control room chiller room and the control room heating, ventilation, and air conditioning (HVAC) room, to be open and unlatched. The individual closed and latched door 70DOR-n-SOO-S and initiated CR-JAF-2O10-08617. ln addition, the engineer documented within the condition report that the door seemed to require more force to shut than normal. After performing the daily review of condition reports the following day, on December 30,2010, the inspectors performed a walkdown and identified door 70bOR-A-300-5 unlatched and slightly ajar. After discovering the door unlatched and ajar, the inspectors verified no Entergy personnelwere in the two rooms associated with door 70DOR-A-300-5, closed and latched the door, and notified control room personnel of the deficiency.
Entergy personnel screened CR-JAF-2010-08617 to the significance level 'D -.Actions Takeri'-and closed the CR on January 13,2011, with closure remarks stating that door 70DOR-A-300-5 was secured closed and the Shift Manager was notified. The inspectors concluded it would have been reasonable to more thoroughly investigate the physical condition of the door at that time, given that there were two separate instances in which door 70DOR-A-300-5 had been identified as unlatched.
On January 21,2011, the inspectors identified door 70DOR-A-300-5 unlatched a third time and noted that the door would, although visually appearing to be fully latched, intermittently fail to automatically latch and require more manual force to close than normal. Eniergy personnel inspected door 70DOR-A-300-5 and identified a piece of metal foreign m-aterial protruding from the bottom of the door between laminate layers, which waslntermittentiy catching upon the ramped door sill and interfering with free closure of the door. Entergy personnel determined that the foreign material was most likely a remnant of the door shell that had fallen inside the door when a hole saw was used to bore the holes for the installation of door knobs on May 12,2009, a corrective action associated with a prior NRC finding documented in inspection report 05000333/2009-002.
In order for the control room emergency ventilation air system (CREVAS) subsystems to be considered operable, the CRE boundary must be maintained such that the CRE occupant dose from a large radioactive release does not exceed the calculated dose in the licensing basis consequence analyses for design basis accidents, and that CRE occupants are protected from hazardous chemicals and smoke. Door 70DOR-A-300-5 must be closed and latched in order to maintain these conditions. Although procedure AP-19.18, "Control Room Envelope Habitability Program," allows intermittent opening of the CRE boundary under administrative controls, as permitted by a note included in TS 9.7.3, the door's condition in this case was not controlled and its state was unknown and unreliable.
Entergy's corrective actions included removing the foreign material and initiating condition reports CR-JAF-2O1 0-0861 7 and CR-JAF-2O11 -00407'
Analysis:
There was an NRC-identified performance deficiency in that Entergy personnet did not promptly identify and correct a condition adverse to quality associated with CRE boundary door 70DOR-A-300-5. The finding was more than minor because it was associated with the configuration control and the barrier performance attributes speci1c to the radiological barrier function of the control room. The finding affected the Barrier Integrity corneistone objective to provide reasonable assurance that physical design barriers (fuet cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events.
The finding was evaluated using Inspection Manual Chapter (lMC) 0609, "Significance Determinalion Process," (SDP) Phase I and screened to a SDP Phase 3 review because the finding represented a degradation of the barrier function provided for the control roor toxic atmosphere and smoke, as well as radiological conditio_ns' Based on senior"gai-nst iisk analyst review, the finding was determined to be of very low safety significance (Gieen) because the amount of time the door was unlatched and ajar was timiteO to 24 days. The mitigating actions immediately required by action statement B'1 are required by B.2 to be verified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to ensure CRE occupant exposures to radiological, chemical, and smoke hazards will not exceed limits. ln addition, action B'3 to restolre the CRE boundary to operable status has a required completion time of 90 days. Therefore, considering the allowed outage time of 90 days, the maximum poientialtime of 24 days during which this condition existed, coupled with the low probability of a design basis aCcident during this time period, results in very low safety significance.
The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution within the corrective action program component because Entergy personnel did not completely and accurately identify the degraded condition of the door (P.1(a) per IMC 0310).
Enforcement:
10 CFR 50, Appendix B, Criterion XVl, "Corrective Action," requires, in part, that, "t\4easures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected." Contrary to the above, between December 29,2010, and January 21, 2011, Entergy personnel did not promptly identify and correct a condition adverse to quality associated with a maliunctionof CRE boundary door 70DOR-A-300-5. This resulted in short periods of time where the CRE boundary was inoperable. Entergy took corrective actions to remove interfering foreign material to improve the reliability of the door closure function.
Because this violation was of very low safety significance and was entered into the CAP as CR-JAF-2011-00407, this violation is being treated as an NCV, consistent with the NRC Enforcement Policy. (NCV 05000333/2011002-01: Control Room Envelope lnoperable due to Unlatched Boundary Door)
.2 Annual Sample: Review of Continued Operabilitv of Liqhthouse Hill Substation 115 kV
Offsite Power Line 3 ===
Inspection Scope The inspectors selected CR-JAF-2005-00109 as a problem identification and resolution sample for a detailed follow-up review. This CR documented Entergy personnel's review regarding whether the TRM may not have correctly reflected the plant licensing basis regarding the 1 15 kV systems as of January 11 ,2005. Offsite power to FitzPa_trick is supplied directly by Lighthouse Hill substation 1 15 kV line 3, and indirectly by South Oswego substation 1 15 kV line 1 via Nine Mile Point - FitzPatrick tie line 4. A loss of line 1 or 4 would require the Lighthouse Hill 1 15 kV line 3 to independently power both reserve station service transformers (RSSTs) and their safety and house loads at FitzPatrick. However, under certain grid loading conditions with line 1 or 4 unavailable, line 3 alone may not provide adequate voltage. This CR was initiated by Entergy personnel to determine operability of the Lighthouse Hill 1 15 kV offsite power line, concurrent with the loss of line 1 or 4, to ensure that voltage would stay within the motor starting capability of FitzPatrick's 1 15 kV and 4 kV electric systems. Additionally, Entergy personnel initiated corrective actions to revise the necessary documents to ensure operation of the line was consistent with the TS requirements and to make modifications to FitzPatrick as necessary. This CR was initiated after an issue was identified on January 10,2005 (documented in CR-JAF-2005-00089), which determined that the plant had continued to operate on a single 1 15 kV offsite supply line (line 3) for greater than the TS-allowed maximum of seven days, as documented in NRC inspection ieport 05000333/2004-005. FitzPatrick personnel performed several operability evaluations (CR-JAF-2005-00109) and determined that, through current operator practices, power monitoring actions, and load limitations, FitzPatrick 1 15 kV offsite circuits are capable of supplying all engineered safeguard loads and performing their intended functions in any configuration until long term corrective actions are completed' The inspectors assessed Entergy's problem identification threshold, apparent cause evaluation (ACE), extent of condition reviews, operability evaluations, and the prioritization and timeliness of corrective actions to determine whether Entergy was appropriately identifying, characterizing, and correcting problems associated with the identified issues and whether the planned or completed corrective actions were appropriate to prevent recurrence. Additionally, the inspectors performed walkdowns of accessible portions of the 1 15 kV system and components to assess if abnormal conditions existed. The inspectors also interviewed plant personnel regarding the identified issues and implemented or planned corrective actions. The documents reviewed are listed in the Attachment.
b.
Findinqs and Observations No findings were identified. The inspectors determined that Entergy personnel properly implemerited their CAP regarding the issue. The CR packages were complete and included an ACE, operability evaluations, extent of condition reviews, use of operating experience, and contained implemented and planned corrective actions. Additionally, the elements of the CRs, ACE, and operability evaluations were detailed and thorough.
lmplemented and planned corrective actions by Entergy personnelwere appropriate to minimize the potential of recurrence.
The inspectors noted that corrective actions included revising the TS bases to revise the definition of qualified offsite circuits to include the 1 15 kV lines, revoking the appropriate TRM section for 115 kV lines being inoperable, and initiating CR-JAF-2005-00109 to perform operability evaluations of the 1 15 kV lines ability to supply adequate power to operate safeguards equipment and house loads when the South Oswego substation/Nine Mile Point Unit 1 power source lines 1 or 4 are out-of-service. The inspectors determined the additional corrective actions included performing several operability evaluations and engineering evaluations to ensure the offsite power capabilities without lines 1 or 4 were adequate to supply power for a loss of coolant accident; implementing an operating shift standing order to perform monitoring of the 115 kV transmission system; establishing communication protocols between Nine Mile Point Unit 1, FitzPatrick, and the grid operator when a 115 kV line is out of service; revising the post contingency voltage alarm set-point from 1 12.5 kV to 1 12.0 kV (subsequently revised to 112.8 kV by further analysis in CR-JAF-2010-03421) to maintain adequate voltage at the 4.16 kV emergency buses; and revising operations procedures to operate the 1 15 kV system within the licensing and design bases.
The inspectors concluded that FitzPatrick personnel had conducted adequate operability evaluations, implemented appropriate agreements with grid operators and had adequate operations procedures in place to provide proper monitoring, evaluation, and response for the 1 15 kV offsite power systems as issues occur. The inspectors determined these actions appear adequate to ensure that the 1 15 kV offsite power lines remain capable of performing their safety function under the various scenarios analyzed.
4OA3 Follow-up of Events and Notices of Enforcement Discretion
.1 (Closed) LER 05000333/2010005-00, High Pressure Goolant lnjection System Declared
lnoperable due to Power Supply Degradation On October 23,2010, Entergy personnel identified a condition that could have prevented the fulfillment of the safety function of the HPCI system, a system needed to remove residual heat. Entergy staff noticed an acrid odor within the control room which was determined to be emanating from 23lNV-79, the HPCI instrument power inverter.
Entergy staff also noted that the casing of 23lNV-79 had a spot of discoloration from apparent localized overheating. Based on these indications, Entergy personnel declared the HPCI system inoperable and implemented an engineering change to replace the degraded inverter, a Topaz Electronics Model N250-GWR-125-60-115, with one of new design.
Entergy personnel performed a visual inspection of the removed inverter's internal components and noted heat and smoke damage associated with a transformer internal to the device. Based on the degraded material condition, Entergy staff considered the inverter to not be capabte of performing its safety function at the time of discovery. The degraded inverter was supplied to FitzPatrick in February 1973 and had been refurbished and reinstalled in the plant on October 1Q,2010. According to Electric Power Research lnstitute EL-5036, Volume 2, "Power Transformers," the typical life expectancy of an energized and loaded transformer is between 30 and 40 years, However, transformers are generally considered beyond the usual scope of items replaced during preventive maintenance refurbishments' The significance of the condition was mitigated by the fact that the automatic depressurization,low pressure coolant injection, core spray, and RCIC systems were available. Corrective actions documented in CR-JAF -201 0-07 341 included implementing a design change and replacing 23lNV-79 with a new model and initiating work orders to replace the remaining Topaz Electronics Model N250-GWR-125-60-115 inverters installed in the plant with the new design. No violation of regulatory requirements occurred and no findings were identified. This LER is closed.
4OAO Meetinos, Includinq Exit Exit Meetinq Summarv The inspectors presented the inspection results to Mr. B. Sullivan and other members of Entergy's management at the conclusion of the inspection on April 27,2011. The inspectors asked Entergy personnel whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified by Entergy's personnel.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Enterqv Personnel
- K. Bronson, Site Vice President
- B. Sullivan, General Manager, Plant Operations
- M. Woodby, Director, Engineering
- B. Finn, Director, Nuclear Safety Assurance
- C. Adner, Manager, Operations
- J. LaPlante, Manager, Security
- J. Barnes, Manager, Training and Development
- T. Raymond, Manager, Project Management
- M. Reno, Manager, Maintenance
- C. Brown, Manager, Quality Assurance, Entergy
- P. Cullinan, Manager, Emergency Preparedness
- V. Bacanskas, Manager, Design Engineering
- D. Poulin, Manager, System Engineering
- P. Scanlan, Manager, Programs and Components Engineering
- J. Pechacek, Manager, Licensing
- E. Wolf, Manager, Radiation Protection
LIST OF ITEMS
OPEN, CLOSED, AND DISCUSSED
Opened and Closed
- 05000333/2011002-01 NCV Control Room Envelope Inoperable due to Unlatched Boundary Door
Closed
- 05000333/201 0005-00 LER High Pressure Coolant Injection System Declared Inoperable due to Power Supply Degradation
Discussed
None