IR 05000293/2010002

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IR 05000293-10-002, on 01/01/2010-03/31/2010; Pilgrim Nuclear Power Station; Plant Modifications and Post-Maintenance Testing
ML101241100
Person / Time
Site: Pilgrim
Issue date: 05/04/2010
From: Diane Jackson
NRC/RGN-I/DRP/PB5
To: Bronson K
Entergy Nuclear Operations
Jackson D E, RGN-I/DRP/PB5/610-337-5306
References
IR-10-002
Download: ML101241100 (41)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION

REGION I

475 ALLENDALE ROAD KING OF PRUSSIA, PENNSYLVANIA 19406-1415 May 4. 2010 Mr. Kevin Bronson Site Vice President Entergy NUclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360-5508 SUBJECT: PILGRIM NUCLEAR POWER STATION - NRC INTEGRATED INSPECTION REPORT 05000293/2010002

Dear Mr. Bronson:

On March 31, 2010, the U,S. Nuclear Regulatory CommiSSion (NRC) completed an inspection at your Pilgrim Nuclear Power Station (PNPS). The enclosed inspection report documents the results, which were discussed on April 15, 2010, with you and members of your staff.

The inspection examined activities performed 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.

The report documents two NRC identified findings of very low safety significance (Green).

These two findings were determined to involve violations 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 their very low safety significance and because they were 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, you should provide a response within 30 days of the date of this inspection report. with the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk. Washington DC 20555-0001; with copies to the Regional Administrator, Region I; the Director. Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Senior R~~sident Inspector at PNPS. In addition, if you disagree with the characterization of any finding in this report, you should pravidea 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 NRC Senior Resident Inspector at PNPS. The information you provide will be considered in accordance with Inspection Manual Chapter 0305.

I I In accordance with 10 CFR 2.390 of the NRC's "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 NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.govlreading-rm/adams.html(the Public Electronic Reading Room).

s()ereIY~fJ

~f. f\ O_

Donald E. JaCkS~

Projects Branch 5 '

Division of Reactor Projects Docket No. 50-293, License No. DPR-35

Enclosure:

Inspection Report 05000293/2010002 w/Attachment: Supplemental Information ec w/enel: Distribution via listServ I. In a

REGION I==

Docket No: 50-293 License No: DPR-35 Report No; 05000293/2010002 Licensee: Entergy Nuclear Operations, Inc.

Facility: Pilgrim Nuclear Power Station (PNPS)

Location: 600 Rocky Hill Road Plymouth, MA 02360 Inspection Period: January 1, 2010 through March 31, 2010 Inspectors: M. Schneider, Sr. Resident Inspector, Division of Reactor Projects (DRP)

B. Smith, Resident Inspector, DRP R. Rolph. Health Physicist, Division of Reactor Safety (DRS)

K. Young, Regional Inspector. DRS W. Schmidt, Senior Reactor Analyst, DRS Approved By: Donald E. Jackson. Chief Projects Branch 5 Division of Reactor Projects Enclosure

l

I

SUMMARY OF FINDINGS

IR 05000293/2010002; 01/01/2010~03/31/2010; Pilgrim Nuclear Power Station; Plant

Modifications and Post-Maintenance Testing.

The report covered a three-month period of inspection by resident and region based inspectors.

Two Green findings were identified, which were determined to be non-cited violations (NCVs).

The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process" (SDP). The cross-cutting aspect for the finding was determined IJsing IMC 0310, "Components Within The Cross-Cutting Areas, to dated February 2010. Findings for which the SDP does not apply may be Green or be assigned a severity level after 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 NRC identified a Green non-cited violation of 10 CFR 50, Appendix B,

Criterion XVI, "Corrective Action," for Entergy's failure to promptly correct a condition adverse to quality. Specifically, Entergy did not correct defective material in their "AD Emergency Diesel Generators (EDG) in a prompt manner which led to emergent maintenance and additional unplanned unavailability of the "A" EDG while they replaced cracked snubber valves. Entergy's corrective actions include entering this issue into the corrective action program and replacing the seven remaining snubber valves on their "A" EDG with those of a material properly hardened and not susceptible to the same mode of cracking.

The inspectors determined that the finding was more than minor because the finding was associated with the Equipment Perform!:mce attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone's objective to ensure the availability of systems that respond to initiating events to prevent undesirable consequences (Le., core damage). Specifically, the "A" EDG was unavailable during snubber valve replacements. The inspectors determined the significance of the finding using (MC 0609.04, "Phase 1 -Initial Screening and Characterization of Findings." The finding was determined to be of very low safety significance (Green) because the finding did not result in a loss of system safety function of a single train for greater than its Technical Specifications outage time, and did not screen as potentially risk significant due to external initiating events. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program component, because Entergy did not take corrective actions in a timely manner. Specifically, Pilgrim did not replace the uA" EDG snubber valves in a prompt manner after repeated fuel leaks from cracked snubber valves over the previous two years. P.1(d) (Section 1R19)

Cornerstone: Barrier Integrity

Green.

The NRC identified a Green non-cited violation of 10 CFR 50, Appendix B,

Criterion V, "Instructions, Procedures and Drawings," for Entergy's failure to accomplish procedures prescribed for activities affecting quality. Specifically, Entergy did not implement their operability determination process or their temporary modification process for compensatory measures needed ito maintain the secondary containment operable. Entergy's corrective actions included deSignating the compensatory measures as necessary to maintain operability for both torus troughs and implementation of temporary modifications for the equipment installed in the plant to support t,hese compensatory measures.

The inspectors determined that the finding was more than minor because the finding was associated with the Human Performance attribute of the Barrier Integrity cornerstone, and adversely affected the cornerstone's objective to provide reasonable assurance that physical deSign barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, operations and engineering personnel did not adequately implement operability determination and temporary modification procedures when degraded and/or non-conforming conditions associated with the secondary containment torus troughs were identified. The inspectors determined the significance ofthe finding using IMC 0609.04, "Phase 1 Initial Screening and Characterization of Findings." The finding was determined to be of very low safety significance (Green) because the finding only represented an impact to the radiological barrier function provided by secondary containment and the standby gas treatment system. This finding had a cross-cutting aspect in the area of Human Performance, Work Practices component, because Entergy personnel did not follow procedures. Specifically, Entergy did not implement their operability determination or temporary modification procedures for compensatory measures needed to maintain the secondary containment operable. H.4(b) (S(~ction 1R18)

Other Findings

A violation of very low safety Significance, which was identified by the licensee has been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensee's corrective action program. This violation and corrective action tracking number are listed in Section 40A7 of the report.

REPORT DETAILS

Summary of Plant Status

Pilgrim Nuclear Power Station (PNPS) began the inspection period operating at 100 percent reactor power. On March 10, 2010, operators reduced power to 46 percent for a backwash of the main condenser due to a storm surge the previous week. Pilgrim returned to 100 percent reactor power later the same day. On March, 15,2010 operators reduced power to 93 percent for a control rod pattern adjustment and returned to 100 percent power the same day.

Operators maintained the reactor at or near 100 percent power for the remainder of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection

Main article: IP 71111.01

.1 Impending Storm

a.

Inspection Scope (1 sample)

On the morning of January 25, 2010, a significant winter storm was tracking to impact the Pilgrim plant The inspectors reviewed Entergy's preparations for the high winds expected to accompany the storm. The inspectors reviewed Entergy's severe weather procedures including; operations during severe weather, coastal storm preparation, and high winds procedures. The inspectors performed a tour of the plant grounds and the switchyard to determine if loose debris or other material could become airborne in the presence of high winds and thereby potentially impact safety related equipment. The documents reviewed during this inspection are listed in the Attachment.

b. Findings

No findings of significance were identified.

1R04 Equipment Aliqnment

Main article: IP 71111.04

.1 Partial System Walkdowns

Main article: IP 71111.04Q

a.

Inspection Scope (4 samples)

The inspectors performed four partial system walkdowns during this inspection period.

The inspectors reviewed the documents listed in the Attachment to determine the correct system alignment. The inspectors performed a partial walkdown of each system to determine if the critical portions of the selected systems were correctly aligned in accordance with these procedures and to identify any discrepancies that may have had an effect on operability. The walkdowns included selected control switch and valve position checks, and verification of electrical power to critical components. Finally, the inspectors evaluated other elements, such as material condition, housekeeping, and component labeling. The following systems Wj9re reviewed based on their risk significance for the given plant configuration: .

b. Findings

No findings of significance were identified .

.2 Complete System Walkdowns (71111.048)

a.

lospection Scoee (1 sample)

The inspectors completed a detailed review of the "B" Residual Heat Removal (RHR)system to assess the functional capability of the system. The inspectors performed a walkdown of the system to determine whether the critical components, such as valves, breakers, and control switches, were aligned in accordance with operating procedures and to identify any discrepancies that could have an effect on operability. The inspectors discussed system health with the system engineer and performed a review of outstanding maintenance work orders to determine whether the deficiencies Significantly affected the "B" RHR system function. The inspectors also reviewed recent condition reports to determine whether "B" RHR equipment problems were being identified and appropriately resolved. The documents reviewed during this inspection are listed in the

.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

Main article: IP 71111.05

Fire Protection - Tours (71111.05Q)

a. Inspection Scope

(5 samples)

The inspectors performed walkdowns of five fire protection areas during the inspection period. The inspectors reviewed Entergy's fire protection program to determine the specified fire protection design features, fire area boundaries, and combustible loading requirements for the selected areas. The inspectors walked down these areas to assess Entergy's control of transient combustible material and ignition sources. In addition, the inspectors evaluated the material condition and operational status of fire detection and suppression capabilities and fire barriers. The inspectors then compared the existing condition of the areas to the fire protection program requirements to determine whether all program requirements were met. The documents reviewed during this inspection are listed in the Attachment. The fire protection areas reviewed were:

  • Fire Area 1.9, Fire Zone 1.8, CRD Pump Quadrant Mezzanine Level;
  • Fire Area 1.9, Fire Zone 1.16, Open Area - North Side of 91 foot elevation; and

b. Finding§ No findings of significance were identified.

1R06 Flood Protection Measures

Main article: IP 71111.06

Internal Flooding Inspection

a. Inspection Scope

(1 sample)

The inspectors walked down the "S" Residual Heat Removal Quadrant, and aSSOCiated flood propagation pathways, to assess the effectiveness of Entergy's internal flood control measures. The inspectors assessed the condition of floor drains, walls, and doors. The inspectors also evaluated whether potential sources of internal flooding were analyzed.

b.

Findin9..s No Findings of Significance were identified.

1R07 Heat Sink Performance

Main article: IP 71111.07

a.

Inspection Scope (1 sample)

The inspectors reviewed one sample of Entergy's program for maintenance, testing, and monitoring of risk significant heat exchangers (HXs) to assess the capability of the HXs to perform their design functions. The inspectors assessed whether the HX program conformed to Entergy's commitments at Pilgrim related to NRC Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment." In addition, the inspectors evaluated whether potential common cause heat sink performance problems could affect multiple HXs in mitigating systems or result in an initiating event. Based on risk significance and prior inspection history, the "A" Residual Heat Removal Heat Exchanger was selected for detailed review by the inspectors.

b. Findings

No findings of significance were identified.

1R11 licensed Operator Regualification Program (7'1111.11)

Resident Inspector Quarterly Review (71111.11 Q) a.

Inspection Scope (1 sample)

The inspectors observed licensed operator performance during an emergency planning drill on February 24, 2010. The inspectors observed crew response to a hostile action based scenario which included a loss of all service water. The inspectors assessed the licensed operators' performance to determine if the training evaluators adequately addressed observed deficiencies. The inspectors reviewed the applicable training objectives from the scenario to determine if they had been achieved. In addition, the inspectors performed a simulator fidelity review to determine if the arrangement of the simulator instrumentation, controls, and tagging closely paralleled that of the control room.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

Main article: IP 71111.12Q

a.

Inspection Scope (3 samples)

The inspectors reviewed the three samples listed below for items such as: (1)appropriate work practices;

(2) identifying and addressing common cause failures; (3)scoping in accordance with 10 CFR 50.65 paragraph
(b) of the Maintenance Rule; (4)characterizing reliability issues for performance;
(5) trending key parameters for condition monitoring;
(6) charging unavailabilitlf for performance;
(7) classification and reclassification in accordance with 10 CFR 50.65 paragraph (a)(1) or (a)(2); and (8)appropriateness of performance criteria for structures, systems, and components (SSCs)/functions classified as paragraph (a)(2) and/or appropriateness and adequacy of goals and corrective actions for SSCs/functions classified as paragraph (a)(1). The documents reviewed during this inspection are listed in the Attachment. Items reviewed included the following:
  • Functional Failure Determination for Reactor Building Closed Cooling Water broken bolt on suction header.

b. Findings

No findings of significance were identified.

1R 13 Maintenance Risk Assessments and Emergent Work Control (71111.13)a.

Inspection Scope (6 samples)

The inspectors evaluated six maintenance risk. assessments for planned and emergent maintenance activities. The inspectors reviewed maintenance risk evaluations, work schedules, and control room logs to determine if concurrent maintenance or surveillance activities adversely affected the plant risk already incurred with out-of-service components. The inspectors evaluated whether Entergy took the necessary steps to control work activities, minimized the probability of initiating events, and maintained the functional capability of mitigating systems. The inspectors assessed Entergy's risk management actions during plant walkdowns. The inspectors reviewed the conduct and adequacy of maintenance risk assessments for the following maintenance and testing activities: .

  • Planned Yellow Risk for Load Shed Testing with the Turbine Auxiliary Oil Pump Out of Service;
  • Emergent Green Risk for Inoperable Reactor Building Closed Cooling Water Train;
  • Planned Yellow Risk for the Testing of Breaker 504 to the Startup Transformer.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

Main article: IP 71111.15

a.

Inspection Scope (6 samples)

The inspectors reviewed six operability determinations associated with degraded or non-conforming conditions to determine if the tJperability determination was justified and if the mitigating systems or barriers remained

.3 vailable such that no unrecognized

increase in risk had occurred. The inspectors also reviewed compensatory measures to determine if the compensatory measures were in place and were appropriately controlled. The inspectors reviewed Entergy's performance against related Technical SpeCifications and Updated Final Safety Analysis Report requirements. The documents reviewed during this inspection are listed in thE;) Attachment. The inspectors reviewed the following degraded or non-conforming conditions:

  • CR-PNP-2010-0229, Rod Block Monitor "A" received Rod Block due to loss of input signal;

b. Findings

No findings of significance were identified.

1R18 Plant Modifications

Main article: IP 71111.18

.1 Temporary Modification Review of Torus Trough Level Indication Installed to SUQPort

Compensatory Measures to Maintain Secondary Containment Operability a.

Inspection Scope (1 sample)

The inspectors reviewed the installation of a temporary torus trough level Indication which had been installed to support secondary containment operability. This temporary level indication was implemented in place of s~{stem level switches which were determined not to meet design basis requirements for secondary containment operability. The inspectors reviewed condition reports, operability evaluations, and the temporary modification procedure to determine if the system modification should be considered a temporary modification and managed as such. In addition, the inspectors reviewed the system modification and design basis documents to ensure the secondary containment function was not adversely affected.

b. Findings

Introduction:

The NRC identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings," for Entergy's failure to accomplish activities affecting quality in accordance with prescribed procedures. Specifically, Entergy did not implement their operability det*ermination process nor their temporary modification process for compensatory measures needed to maintain the secondary containment operable.

Description:

On December 22, 2009, a design engineer conducting a walkdown of the torus room noted that one of the torus troughs (the "A" reactor auxiliary bay floor sump trough) was dry and that the other trough was low in water level. The torus troughs are deSigned such that reactor auxiliary bay floor sump piping that penetrates the secondary containment are directed to the torus troughs and covered with water to provide a water seal and ensure secondary containment integrity. The engineer notified the control room and operators entered Technical Specification (TS) 3.7.C, Secondary Containment, refilled both torus troughs, and exited the TS. Entergy determined that a low water level switch in the "A" trough had malfunctioned resulting in a failure to receive an alarm in the control room. When this low level alarm is received in the control room, operators are directed by procedures to refill the affected trough in order to maintain secondary containment integrity. On December 30, 2009, in response to the inoperable level switch, Entergy implemented compensatory measures to install a remote camera to monitor the water level in the uN torus trough until the level sWitch could be repaired.

On December 31, 2009, the inspectors reviewed the Entergy's corrective actions for the torus trough issue. The inspectors noted that while an operator could observe water in the torus trough using the remote camera, there was no level indicating device to identify when the trough water level was too low to provide a seal and ensure secondary containment integrity. Operators subsequently installed level indicating devices (ruler strapped to the reactor auxiliary bay sump piping); however, they did not designate these compensatory measures as being specified to maintain operability. Requirements associated with compensatory measures to maintain operability are discussed in EN OP-104, Revision 4, "Operability Determination Process," in Sections 5.4(1), (9). and (10), Section 5.5(6}, and Section 5.6. These sections describe the need to identify when compensatory measures are required to maintain operability, to review for the applicability of other processes, such as the temporary modification process, that may be affected, to assess whether the compensatory measures can impact other plant equipment or procedures, and the need to periodically review these compensatory measures to maintain awareness and to ensure timely corrective actions. The inspectors questioned why the torus trough m<>nitoring was not deSignated as a compensatory measure to maintain operability; however, the "A" torus trough level switch was subsequently repaired on January 21,2010, and available to warn operators of lowering trough water level. Operators, however, continued to monitor the 'A' torus trough water level using the remote camera and ruler.

On January 27, 2010, design engineering determined that the existing torus trough low water level switch setpoint was not adequate to ensure that secondary containment design requirements were met Given the conclusion by design engineering, the inspectors questioned why operators had not consequently designated the actions to observe torus trough water level as compensatory measures to maintain operability of secondary containment and whether these actions would now apply to both torus troughs. On March 6, 2010, operators designated these compensatory measures as necessary to maintain secondary containment operability. The inspectors then questioned operators and design engineering about whether the equipment installed to support the compensatory measures should be designated as temporary modifications.

EN-DC-136, Revision 5, "Temporary Modifications," Attachment 9.2, states, in part, that "specific temporary physical plant alterations specified for compensatory measures to maintain operability would be a Temporary Modification." On March 25, 2010, system engineering established a corrective action to issue temporary modifications for both torus troughs.

Analysis:

The inspectors determined that Ente!rgy's inadequate implementation of their operability determination process (EN-OP-1 04) and their temporary modification process (EN-DC-136) for compensatory measures that were required to maintain operability of secondary containment was a performance deficiency. Traditional enforcement did not apply, as the issue did not have actual or potential safety consequence, had no willful aspects, nor did it impact the NRC's ability to perform its regulatory function. A review of NRC Inspection Manual Chapter (IMC) 061':;~, Appendix E, "Minor Examples," revealed that no minor examples were applicable to this finding. The inspectors determined that the finding was more than minor because the finding was associated with the Human Performance Attribute of the Barrier Integrity Cornerstone, and adversely affected the cornerstone's objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, operations and engineering personnel did not implement operability determination and temporary modification procedures when degraded and/or non conforming conditions associated with the secondary containment torus troughs were identified. The inspectors determined the significance of the finding using IMC 0609.04, "Phase 1 -Initial Screening and Characterization of Findings." The finding was determined to be of very low safety significance (Green) because the finding only represented an impact to the radiological barril~r function provided by secondary containment and the standby gas treatment system.

This finding had a cross-cutting aspect in the area of Human Performance, Work Practices component, because Entergy personnel did not follow procedures.

Specifically, Entergy did not implement their operability determination or temporary modification procedures for compensatory measures needed to maintain the secondary containment operable. H.4{b)

Enforcement:

10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings." requires, in part, that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures. Contrary to the above, Entergy did not accomplish the requirements outlined in procedures for the determination of operability or for the identification of temporary modifications when compensatory measures were identified which were necessary to maintain secondary containment operability. Specifically, between December 30,2009 and March 6,2010, Entergy did not adequately implement EN-OP-104 and designate installed remote cameras and level indicating devices as compensatory measures that were required to ensure adequate water level was maintained to keep the secondary containment water seal operable.

Additionally, between March 6 and March 25,2010, Entergy did not adequately implement EN-OC-136 to designate the temporary physical plant alterations (remote cameras and level devices) specified as compensatory measures to maintain operability as Temporary Modifications. Entergy's corrective actions included deSignating the compensatory measures as necessary to maintain operability for both torus troughs and implementation of temporary modifications for the equipment installed in the plant to support these compensatory measures. The secondary containment torus trough issues are documented in CR-PNP-2009-5295, CR-PNP-2009-5309, and CR-PNP-2010-0014.

Because this finding is of very low safety signilicance and Entergy has entered it into their corrective action program, this violation is being treated as an NCV, consistent with Section VI.A.1 ofthe NRC Enforcement Policy. NCV 0500029312010002-01, Failure to Implement Operability Determination and Temporary Modification Processes for Compensatory Measures Required to Maintain Operability of Secondary Containment.

.2 Temporary Modification to Provide 24VDC Power during "A" Battery Testing

a.

Inspection Scope (1 sample)

The inspectors reviewed Temporary Modification EC12349, uProvide 24VDC Power during "A" Battery Testing," to determine whether the performance capability of the "An 24VDC safety related bus had been degraded through the modification. The inspectors reviewed Control Room and procedural drawings. relevant condition reports. and work orders to ensure the temporary modification did not adversely affect the 24VDC system.

The inspectors reviewed the annotated drawings to determine whether they properly reflected the temporary modification. The inspectors also walked down the battery and switchgear rooms to ensure tagging was appropriate for the modification.

b. Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing

Main article: IP 71111.19

a.

Insl2ection Scope (7 samples)

The inspectors reviewed seven samples of post-maintenance tests (PMT) during this inspection period. The inspectors reviewed these activities to determine whether the PMT adequately demonstrated that the safety-related function of the equipment was satisfied. given the scope of the work performed, and that operability of the system was restored. In addition, the inspectors evaluated the applicable test acceptance criteria to verify consistency with the associated design and licensing bases, as well as Technical Specification requirements. The inspectors also evaluated whether conditions adverse to quality were entered into the corrective action program for resolution. The documents reviewed during this inspection are listed in the Attachment. The following maintenance activities and their post-maintenance tests were evaluated:

  • C-19A Electronics are unresponsive;
  • Standby Gas Treatment "A" Train Backdraft and Outlet Damper maintenance and testing;
  • HPCI Electrical maintenance including HPGI Condensate Pump Motor Brush replacement, HPCI Auxiliary Lube Oil Pump Motor Brush replacement, and HPCI Gland Seal Condenser Blower Pump Brush replacement; n
  • HPCI replacement of Rupture Disk, repair of Temperature Control Valve (TCV-2301 230) and additional mechanical maintenance postwork tests; and

b. Findings

Introduction:

The NRC identified a Green non~cited violation of 10 CFR 50, Appendix B, Criterion XVI, "Ccmectfve Action." for Entergy's failure to promptly correct a condition adverse to quality. Specifically. Entergy did not correct leaking snubber injection valves on the "A" Emergency Diesel Generator (EDG) in a timely manner.

Description:

On March 12,2008, Pilgrim's "A" EOG exhibited a fuel oil leak from its 7R cylinder during the monthly surveillance. The fuel leak was determined to be a symptom of a cracking phenomenon of the fuel injector snubber valve. The snubber valve serves to dampen pulsations from the positive displacement injector pump and to act to keep the fuel tube full on the back stroke of the pump. The "A" EOG was removed from service to replace the snubber valve on the 7R cylinder and was subsequently returned to service. A 10 CFR 21 report from Entergy's Palisades plant was written on April 2, 2008 and listed Pilgrim as a plant susceptible to this cracking snubber valve phenomenon. Specifically, the particular material used in these snubber valves was susceptible to material defects from improper through-hardening during the manufacturing process. Entergy then discovered additional cracked snubber valves on its "A" EOG 9R cylinder on June 9, 2008, during a maintenance overhaul window and then on its "A" EDG 6R cylinder during subsequent post work testing on June 14, 2008.

CR-PNP-2008-1894 and CR-PNP-2008-1952 were written. After three of the 18 snubber valves were replaced, Entergy conducted an extent of condition review and determined that nine of the remaining 15 snubber valves on the ~A" EDG would be susceptible to cracking due to material defects. The susceptible snubber valves on the "B" EOG previously had been replaced during an overhaul in 2009. The apparent cause recommended replacement of these snubber valves before the next scheduled overhaul in the summer of 2010. This activity did not take place.

On March 10, 2010, Entergy discovered the "A" EOG 2L and 3L fuel cylinders leaking from cracked snubber valves. Entergy removed the "A" EDG from service for emergent maintenance and replaced the affected snubber valves. Their action plan is to replace the remaining seven susceptible snubber valves in June 2010, during the next planned "AI> EDG overhaul.

Analysis:

The performance deficiency was that Entergy did not promptly correct a condition adverse to quality, cracked snubber injection valves on their "An EDG. The failure to correct this condition in a timely manner (over two years from identifying the first leaking cylinder) resulted in additional unplanned unavailability for the "AU EOG.

Traditional enforcement did not apply; as the issue did not have actual safety consequence, had no willful aspects, nor did it impact the NRC's ability to perform its regulatory function. A review of NRC Inspection Manual Chapter (IMC) 0612, Appendix E, "Minor Examples," revealed that no minor examples were applicable to this finding.

The inspectors determined that the finding was more than minor because the finding was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone's objective to ensure the availability of systems that respond to initiating events to prevent undesirable consequences (Le .* core damage). Specifically, unplanned maintenance added additional unavailability to the "A" EDG during snubber valve replacement. The inspectors determined the significance of the finding using IMC 0609.04, "Phase 1 Initial Screening and Characterization of Findings." The finding was determined to be of very low safety significance (Green) because, although additional "N' EDG unavailability was incurred, the finding did not result in a loss of system safety function of a single train for greater than its Technical Specifications allowed outage time and did not screen as potentially risk significant due to external initiating events.

This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program component, because Entergy did not take corrective actions in a timely manner. Specifically, Pilgrim did not replace the "A" EDG snubber valves in a prompt manner after repeated fuel leaks from cracked snubber valves over the previous two years. P.1(d}

Enforcement:

10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," requires in part that measures shall be established to assure that conditions adverse to quality, such as defective material and equipment are promptly identified and corrected.

Contrary to the above, Entergy was not prompt in correcting defective material in their "Aw EDGwhich led to emergent maintenance I::lnd additional unplanned unavailability of the ~N' EDG while they replaced cracked snubber valves. Entergy's corrective actions include replacing the seven remaining snubber valves on their "A" EDG with those of a material properly hardened and not susceptible to the same mode of cracking. Entergy has captured these failures in their corrective action program as CRs 2008-0852, 2008-1071, 2008-1894, 2008-1952, and 2010-0898. Because this finding is of very low safety significance and Entergy has entered it into their corrective action program, this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy. NCV 05000293/2010002-02, Untimely Corrective Actions to Promptly Correct Leaking Snubber Valves on the "A" Emergency Diesel Generator.

1R22 Surveillance Testing

Main article: IP 71111.22

a. Inspection Scope

{6 samples}

The inspectors witnessed six surveillance activities and/or reviewed test data to determine whether the testing adequately demonstrated equipment operational readiness and the ability to perform the intended safety-related functions. The inspectors reviewed selected prerequisites and precautions to determine if they were met and if the tests were performed in accordc:lnce with the procedural steps.

Additionally, the inspectors evaluated the applicable test acceptance criteria for consistency with associated design bases. licensing bases, and Technical Specification requirements. The inspectors also evaluated whether conditions adverse to quality were entered into the corrective action program for resolution. The following surveillance tests were evaluated:

b. Findings

No findings of significance were identified.

Cornerstone: Emergency Preparedness (EP)

1EP6 Drill Evaluation

Main article: IP 71114.06

a. Inspection Scope

(1 drill observation sample)

The inspectors observed an emergency planning drill on February 24, 2010. The inspectors evaluated the emergency response organization performance in the simulator, in the alternate Technical Support Center, and in the Emergency Operations Facility. for a hostile action based scenario which escalated to a General Emergency.

The inspectors assessed the implementation of Emergency Action Level classification and notification decisions as well as Protective Action Recommendation development and notifications. The inspectors also assessed whether Pilgrim's critique of the exercise assessed all of the drill's observations and findings.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

(RS)

Cornerstone: Occupational and Public RadiatIon Safety

2RS0 5 Radiation Monitoring Instruments

Main article: IP 71124.05

a. Inspection Scope

During the period between February 8 and 12,2010, the inspectors performed the following activities to verify that Entergy was ensuring the accuracy and operability of radiation monitoring instrumentation. Implementation of these controls was reviewed against the criteria contained in 10 CFR 20, relevant Technical Specifications, and Entergy's procedures.

Inspection Planning
  • The inspectors reviewed the Updated Final Safety Analysis Report (UFSAR) to identify radiation instruments associated with monitoring area radiological conditions including airborne radioactivity, process streams, effluents, material/articles, and workers.
  • The inspectors obtained a listing of all survey instrumentation including air samplers, small article monitors (SAMs), personnel contamination monitors (PCMs), and other monitors used to detect internal contamination. The inspectors reviewed the list to determine if an adequate number and type of instruments are available to support operations.
  • The inspectors obtained and reviewed copies of evaluation reports of the radiation monitoring program since the last inspection.
  • The inspectors obtained and reviewed copies of procedures used for instrument source checks and calibrations.
  • The inspectors reviewed area radiation monitor set point values and basis.

Walkdowns and Observations

  • The inspectors toured the Turbine and Reactor buildings and observed the condition of the Steam Jet Air Ejector monitors, the Main Stack Ventilation monitors, the Reactor Building Ventilation monitors, and the Radioactive Waste Discharge monitor.

These monitor configurations aligned with Pilgrim's ODCM descriptions.

  • The inspectors checked the calibration due dates and source check stickers for portable survey instruments ready for issue or in the field. The type of instruments checked included RO-2As, RO-20s, TeleplJles, and Ludlum 3s.
  • The inspectors observed a technician perform instrument $ource checks during the back shift. The inspectors verified that the instrument source checks included exposures at each high-range scale. The source check observations included RO 2s, RO-2As, RO-20s, Telepoles, and Ludlum 35.
  • The inspectors verified Area Radiation Monitors (ARM) and Continuous Air Monitors (CAM) were appropriately positioned relative to the radiation source(s) they were intended to monitor. The inspectors compared the monitor response with actual area conditions for several ARMs.
  • The inspectors observed the daily source checks for PM-7 #600. SAM #308, and Aptec PMW:'2 #52. The inspectors verified the source checks were in accordance with the manufacturer's recommendations and Pilgrim's procedures.

Calibration and Testing Program

Process and Effluent Monitors

  • The inspectors verified for more than four effluent monitor instruments that channel calibration and functional tests were performed consistent with radiological effluent technical specifications. The inspectors also verified that the source calibrations use National Institute of Standards and Technology (NIST) traceable sources or secondary measuring that has been calibrated to !'JIST standard. The inspectors verified that the sources used represent the plant nuclide mix.
  • The inspectors verified that effluent monitor alarm set points are established as provided in the aDCM and station procedures.
  • There were no changes to effluent monitor set-points during this inspection period.

Laboratory Instrumentation

  • The inspectors verified that the daily performance checks and calibration data indicate the frequency of calibration is adequate and there is no degradation of instrument performance.

Whole Body Counter

  • The inspectors reviewed the methods and sources used to perform the Whole Body Counter (WBC) checks prior to daily use. The inspectors verified the checks are appropriate and align with the plant's isotopic mix.
  • The inspectors reviewed the WBC calibration reports completed since the last inspection. The inspectors verified the calibration sources and phantoms used were appropriate and representative of the plant source term.

Post Accident Monitoring Instrumentation

  • The inspectors reviewed the April 18, 2009 calibration records for the Drywell high range monitors, RIT-1001-606A and RIT-1001-606B.
  • The inspectors verified that an electronic calibration for the Drywell high-range monitors was performed and included each decade above 10 rem/hour. The inspectors also verified that a source calibration was performed and included an exposure for at least one decade below 10 rem/hour.
  • The inspectors verified the acceptance criteria were reasonable.
  • The inspectors reviewed the calibration records and availability for the Main Stack Ventilation and Reactor Building Ventilation high range monitors.
  • The inspectors reviewed Pilgrim's capability to collect high-range, post accident iodine effluent samples.
  • There were no opportunities to observe electronic or source calibrations of the high range monitors during this inspection.

PMs. PCMs. and SAMs

  • The inspectors verified that the alarm set point values for PM 7s, SAM s. and PMW-2 are reasonable to ensure licensed material is not released from Pilgrim.
  • The inspectors reviewed the calibration records for PM 7 # 600. SAM # 308, and PMW-2#52.

Portable Survey Instruments. ARMs. Electronic Dosimetry, and Air Samplers/CAMs

  • The inspectors reviewed calibration records for ARMs, an AMS-4, an RO-20, an RO 2, an RO-2A, a Radeco, a PM-7, a Radiation Air Sampler (RAS) Flow Gauge, a Ludlum 3, and a Ludlum 177. The inspectors reviewed the detector measurement geometry and calibration methods for ARMs and portable radiation survey instruments. The inspectors had a technician demonstrate the use of the instrument calibrator.
  • There were no opportunities to review the corrective actions taken for instruments found significantly out of calibration during this inspection.

Instrument Calibrator

  • The inspectors reviewed the current output tables for Entergy's portable survey and ARM instrument calibrator unit. The inspectors verified that Entergy periodically measures the calibrator output over the range of the instruments.
  • The inspectors verified the calibrator is sent for periodic calibration to a facility that uses NIST traceable sources.

. Calibration and Check Sources

  • The inspectors reviewed Entergy's 10 CFR 61 source term to verify that the calibration sources used are representative of the types and energies of radiation encountered in the plant.

Problem Identification and Resolution

  • The inspectors reviewed thirteen
(13) cond!ition reports related to radiation monitoring instrumentation and verified that appropriate corrective actions have been taken or initiated. The inspectors verified that problems are being identified at the appropriate threshold and are properly addressed for resolution.

b. Findings

No findings of significance were identified.

OTHER ACTIVITIES

lOA]

40A1 Performance Indicator (PI) Verification (71151) Cornerstones: Mitigating Systems and Barrier Integrity a.

Inspection Scope (3 samples)

The inspectors reviewed PI data to determine the accuracy and completeness of the reported data. The review was accomplished by comparing reported PI data to confirmatory plant records and data available in plant logs, Licensee Event Reports (LER), Condition Reports (CRs), and NRC inspection reports. The acceptance criteria used for the review was Nuclear Energy Institute (NEI) 99-02, Revision 6, "Regulatory Assessment Performance Indicator Guidelines" and NUREG-1022, Revision 2, "Event Report Guidelines 10CFR 50.72 and 50.73." The documents reviewed during the inspection are listed in the Attachment. The following performance indicators were reviewed:

  • Mitigating System Cornerstone, Safety System Functional Failures from the first quarter of 2009 through the fourth quarter of 2009; I

I

!

quarter of 2009 through the fourth quarter of 2009; and

b. Findings

No findings of significance were identified.

40A2 Identification and Resolution of Problems (71152)

.1 Review of Items Entered into the Corrective Action Program (CAP)

a. Inspection Scope

The inspectors performed a screening of each item entered into Entergy's CAP. This review was accomplished by reviewing printouts of each Condition Report (CR),attending daily screening meetings and/or accessing Entergy's database. The purpose of this review wasto 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: Review of an Automatic Scram Resulting from a Switchyard Breaker

Fault During a Severe Winter Storm, and Momentary Loss of all 345 kV Off-Site Power to the Startup Transformer from a Switchyard Breaker Fault a.

Inspection Scope (1 sample)

The inspectors selected condition reports (CR) PNP-2008-03962 and PNP-2008-039BO as problem identification and resolution (PI&R) samples for a detailed follow-up review.

CR PNP-200B-03962 documented an automatic scram due to a switchyard fault during a severe winter storm on December 19, 200B. CR PNP~2008-03980 documented the momentary loss of all 345 kV off-site power to the startup transformer (SUT), X4, while the plant was in hot shutdown on December 20, 2008. Entergy determined that the cause of the automatic scram on December 19, 2008, was conductive snow/ice buildup on the non-conductive porcelain surfaces ofthe ACS-105 circuit breaker bushing during a severe snow storm. The snowlice accumulation on the circuit breaker "A" phase bushing resulted in an electrical fault causing a reactor scram with the plant operating at 100% power. Entergy determined that the cause of the momentary loss of all 345 kV off-site power to the SUT was a phase "B" to ground fault on the switchyard line 355 bus section. A directional ground overcurrent relay (DGOR) at the Auburn Street Station facility was incorrectly set and an incorrect signal was sent to trip a breaker (ACB-103)for the 342 off-site power line in the Pilgrim Nuclear Power Station (PNPS) switchyard.

This event was initiated by an electrical fault caused by accumulated snow falling from the overhead line 355 bus section and bridging the gap to the "S" phase arc horn. This caused the momentary loss of the 355 line and the 342 line. By design, the 342 line should not have tripped. The 342 line tripped due to an incorrect overcurrent setting of the DGOR at the Auburn Street Station facility. The DGOR was incorrectly set because of an error in a grid computer model used by the Auburn Street Station facility owner to determine the proper setting for the DGOR. As a result, proper clearing of PNPS switchyard faults in the 345 kV switchyard for the 342 line did not occur. The proper setting of the DGOR would have prevented thE~ 342 line from tripping thus maintaining an off-site power source to PNPS.

The inspectors assessed Entergy's problem identification threshold, cause analyses, extent of condition reviews, operability determinations, and the prioritization and timeliness of corrective actions to determine whether Entergy was appropriately identifying, characterizing, and correcting problems associated with these issues and whether the planned or completed corrective 81ctions were appropriate to prevent recurrence. Additiona"y, the inspectors performed walkdowns of the PNPS 345 kV switchyard to assess if abnormal conditions existed. The inspectors also interviewed cognizant plant personnel regarding the identified issues and implemented corrective actions. Specific documents reviewed are listed in the attachment to this report.

b. Findings and Observations

No findings of significance were identified.

The inspectors determined that Entergy properly implemented their corrective action process regarding the initial discovery of the reviewed issues. The CR packages were complete and included root cause evaluations (RCE), operability determinations, extent of condition reviews. use of operating experience, and corrective actions. Additionally, the elements of the condition reports and RCEs were detailed and thorough. Corrective actions appeared appropriate to minimize the potential of flashover faults in the 345 kV switchyard and prevent recurrence of the momentary loss of both 345 kVoff-site power lines. The inspectors determined that corrective actions for the December 19,2008, event included revising severe weather procedures to provide enhanced monitoring of the PNPS 345 kV switchyard/components during severe snow weather events and included implementation of specific operator actions as a result of degrading conditions (snow/ice buildup) in the switchyard. The corrective actions for the December 20,2008, event included revising the set-point for the Auburn Street Station facility DGOR for the 342 off-site power line to operate at the appropriate fault current setting. The new setting will allow the 342 offsite power line to remain energized for a fault on the 355 off site power line, thus maintaining one off-site electrical power source. PNPS coordinated with the interconnection transmission owner, the Auburn Street Station facility owner, and the transmission operator to verify and validate the new fault current setting for the Auburn Street Station DGOR. Additionally, it appeared that Entergy took appropriate corrective actions and post-trip reviews to evaluate and replace damaged switch yard components prior to placing them back into service following the reviewed events.

.3 Annual Sample: Review of Reactor Core Isolation Cooling (RCIC) Discharge and

Suction Pressurization After Shutdown From Routine Testing a.

Inspection Scope (1 Sample)

The inspectors reviewed the circumstances leading to the pressurization of the RCIC discharge and suction piping after a January 6,2010, routine surveillance testing and the actions taken by Entergy to evaluate this condition. Specifically, following quarterly surveillance testing, after the pump was secured, pump discharge pressure slowly rose from the normal pressure of 33 psig to at or near the suction relief valve setpoint of 100 psig over approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. To relieve the pressure, operators opened the pump minimum flow valve from the control room three times, venting the pressure to the torus.

Following the third venting the suction pressune did not increase.

The safety concern was the potential for back leakage of the significantly higher temperature (approximately 365 QF) feedwater causing vapor voids in the RCIC pump discharge and suction piping, which could cause water hammer during pump start or damage to the pump, respectively following a pump start. Continuous leakage of feedwater, driven by an approximate 1200 psig, past the RCIC injection check valve, injection double disc gate valve and the CST suction check valve could increase the temperature of the water in the discharge and suction piping above the saturation temperature for the static pressure conditions.

b. Findings

/Observations No findings of significance were identified.

The inspectors reviewed the following and found Entergy's actions adequate to address the condition:

  • Condition Report CR*PNP~201 0-00063, Received RCtC pump Suction Pressure High alarm, dated January 6,2010;
  • Process Applicability Determination, dated January 12, 2010, used to justify revision 14 to the RCIC high suction pressure alarm response procedure (ARP-C904L-A3), to specify opening the pump minimum flow valve to reduce the pressure;
  • Operability/Functionality Evaluation, dated January 19,2010, which determined that RCtC was operable and that either a small amount of leakage past the injection check valve and normally closed injection double disc gate valve or thermal expansion of initially cold water from the CST caused the pressurization. This included verification, using a work order on January 7,2010, that the system piping up to the normally closed MOV-49 was full of water using ultrasonic measurement and verification that the piping was at ambient temperature of approximately 80 OF.

The speculation was that the downstream disc of MOV-49, which was closed in this test, had been pushed just off its closed seat by the pump's discharge pressure, allowing a small amount of leakage which pressurized the suction side of the pump; and,

  • Troubleshooting was conducted during the next quarterly RCIC surveillance test on March 3,2010, which indicated that the cause was leakage past MOV-49 versus thermal expansion of colder CST water. Specifically this test closed MOV-48 upstream of MOV-49 following the ST and monitored the pressure between the two valves and the pump suction pressure. Pressure between MOV-49 and MOV-48 increased slowly to just over 190 psig in about three and a half hours; while the pump suction pressure did not change. This indicated that leakage past MOV*49 was the cause of the pressurization.

Based on the above, the inspectors concluded that given the documented condition post-surveillance test leakage past MOV-49 was relatively small and not an operability concern relative to discharge or suction piping voiding. The inspectors also concluded that venting the piping, as directed by the alarm response procedure for high RCIC suction pressure, using the minimum flow valv,e was an acceptable contingency action.

40A3 Event Follow-up (71153)

.1 Operator Performance During Condenser Backwash

a.

Inspection Scope (1 sample)

The inspectors observed an infrequently performed evolution on March 10, 2010.

Specifically, the inspectors observed a plant dl)wnpower to support backwashing of the condenser. The inspectors observed the operators reduce power from 100 percent to 46 percent by lowering recirculation flow and inserting control rods. The inspectors reviewed procedural guidance and the power maneuver plan, and observed control room conduct and control of the evolutions. Tine documents reviewed during this inspection are listed in the Attachment b.

Finding~

No findings of significance were identified .

.2 Loss of Standby Gas Treatment due to Demister Door Open

a.

Inspection Scope (1 sample)

On March 25,2010, an Entergy security officer during rounds discovered an open demister door to charcoal vaults on the "B" Standby Gas Treatment (SBGT) System.

The security officer notified the control room and operations declared both trains of SBGT inoperable. Operators entered Technical Specification 3.7.8, whtch requires Standby Gas to be restored within a 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> timeframe. Operations dispatched an operator, closed the demister door, exited the Technical Specification, and made notifications for the loss of SBGT. The inspectors reviewed control room logs, Technical SpeCifications, and notification reqUirements.

b. Findings

See Section 40A7.

.3 (Closed) Licensee Event Report (LER 05000293/2009-001-00), Target Rock Relief Valves

Test Pressure Exceeded Limit Due to Setpoint Variance a.

Inspection Scope (1 sample)

On June 15, 2009, Entergy identified that three out of four target rock relief valve pilot assemblies exceeded their Technical Specification pressure limits during routine testing post Refueling Outage 17. Entergy had replaced all four of their pilot assemblies for their main steam relief valves during Refueling Outage 17. NRC review of upward pressure setpoint drift is documented in Regulatory Issue Summary 2000-12, Resolution of Generic Issue 165, "Spring Actuated Safety and Relief Valve Reliability." Additionally, specific safety relief valve issues at Pilgrim are likewise documented in II~ 05000293-2007-06, and in the problem identification and resolution section of IR 05000293-2008-005. No new findings or violations of significance were identified during the inspector's review. The LER provided an accurate description of planned follow-up actions related to Pilgrim's safety relief valves. This LER is closed.

b. Findings

No findings of Significance were identified .

.4 (Closed) Licensee Event Report (LER 05000293/2009-a02-00), Failure to Meet Technical

Specification Requirements for Secondary Containment a.

Inspection Scope (1 sample)

The inspectors reviewed Entergy's actions associated with LER 05000293/2009-002-00, which are addressed in the CAP as CR-PNP-2009-5295 and CR-PNP-2009-5309. The event was discussed in NRC Inspection Report (IR)05000293/2009005 and related inspection findings are discussed in Sections 1R18 and 40A7 of this report. The documents reviewed during the inspection are listed in the Attachment. This LER is Closed.

b. Findings

See Sections 1R18 and 40A7.

40A6 Meetings. Including Exit On February 11, 2010,the inspectors performed a Radiation Safety exit meeting with the plant at 2:00 P.M. Kevin Bronson, Site Vice President, attended the meeting. The inspectors verified that no proprietary information was provided to the inspectors during the inspection.

On March 18, 2010, the inspectors presented a debrief of the inspection results to Mr. Stephen Beneduci, Engineering Supervisor, and Mr. Jeffrey Keene, Systems Engineer. This inspection report feeder does not contain proprietary information.

On April 21, 2010, the resident inspectors conducted an exit meeting and presented the preliminary inspection results to Mr. Kevin Bronson, and' other members of the Pilgrim staff. The inspectors confirmed that proprietary information provided or examined during the inspection was controlled and/or returned to Entergy and the content of this report includes no proprietary information.

40A7 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 NCVs.

.

  • Technical Specification (TS) 5.4.1 requires written procedures shall be established, implemented, and maintained covering procedures specified in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Contrary to this, on March 25, 2010, a Demister Door on the "B" train of the Standby Gas Treatment (SBGT), required to be closed following the surveillance activity in procedure 8.M.3-18, was found to be left open by an Entergy security officer conducting normal rounds. SBGT was declared inoperable and then was restored to service. This event is documented in Entergy's corrective action program as CR-PNP-201 0-1079. The finding is of very low safety significance because the finding only represents a clegradation of the radiological barrier function provided for the SBGT system.

. radiological barrier function provided by secondary containment and the standby gas treatment system.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Entergy personnel:

S. Beneduci Engineering Supervisor

S. Bethay Director. Nuclear Safety Assurance

K. Bronson Site Vice President

B. Byrne Licensing Engineer

S. Das

R. Hargat

G. Jennings

Electrical Design Engineer

Radiation Protection Technician

Radiation Protection Technician

I

K. Kampschneider Senior Systems Engineer  !

J.Keene Systems Engineer

/.

W.Lobo Licensing Engineer i

J. Martin Electrical Maintenance Superintendent

M. McDonnell Operations Assistant Manager

T. McElhinney Chemistry Manager

D. Noyes Operations Manager

J. Priest Radiation Protection Manager

K. Sejkora Senior Chemist

R. Smith General Manager Pilgrim Operations

M. Thornhill Radiation Protection Supervisor

LIST OF ITEMS

OPENED, CLOSED AND DISCUSSED

Opened and Closed

NCV

05000293/2010002-01 Failure to Implement Operability Determination Process and Temporary Modilication Process for Compensatory Measures Required to Maintain Operability of Secondary Containment.

NCV

05000293/2010002-02 Inadequate Corrective Actions to Promptly Correct Leaking Snubber Valves on the "A" Emergency Diesel Generator.

Closed

LER

05000293/2009-001-000 Target Rock Relief Valves Test Pressure Exceeded Limit Due to Setpoint Variance LER
05000293/2009-002-000 Failure to Meet Technical Specification Requirements for Secondary Containment

LIST OF DOCUMENTS REVIEWED

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