IR 05000293/2008005
Download: ML090370022
Text
February 5, 2009
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 05000293/2008005 - EXERCISE OF ENFORCEMENT DISCRETION
Dear Mr. Bronson:
On December 31, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Pilgrim Nuclear Power Station (PNPS). The enclosed report documents the results, which were discussed on January 7, 2009, with you and 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.
The report documents one NRC-identified finding, and one self-revealing finding of very low safety significance (Green). Both of these findings were determined to involve violations of NRC requirements. 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 (NCV)s, in accordance with Section VI.A.1 of the NRC's Enforcement Policy. If you contest the NCVs in this report, 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 Resident Inspector at the Pilgrim Nuclear Power Station.
Additionally, this report closes one issue involving an incorrect entry into Technical Specification (TS) 4.0.3, Surveillance Requirement Applicability, after you determined Reactor Protection System (RPS) time response testing had not been conducted on several RPS scram contactors in 2007 (Unresolved Item (URI) 05000293/2007-003-04). The NRC determined that TS 3.1, Reactor Protective System, should have been entered vice entering TS 4.0.3, because the surveillance on this portion of the RPS system had never been performed. Although the incorrect entry into TS 4.0.3 is a violation of NRC requirements, the NRC identified no performance deficiency and that discretion is warranted because: (1) licensee current basis documents do not specifically clarify the distinction between a missed surveillance and one that has never been performed, (2) the licensee subsequently completed the surveillance testing satisfactorily, and (3) the issue was of very low safety significance. Based on these facts, I have been authorized, after consultation with the Director, Office of Enforcement, and the Region I Regional Administrator, to exercise enforcement discretion in accordance with Section VII.B.6 of the Enforcement Policy, and refrain from issuing enforcement action for the violation.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosures, 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.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,/RA/ Original Signed By; David C. Lew, Director Division of Reactor Projects Docket No. 50-293 License No. DPR-35
Enclosure:
Inspection Report 05000293/2008005
w/Attachment:
Supplemental Information
cc w/encl: Vice President, Operations, Entergy Nuclear Operations Vice President, Oversight, Entergy Nuclear Operations Senior Manager, Nuclear Safety & Licensing, Entergy Nuclear Operations Senior Vice President and COO, Entergy Nuclear Operations Assistant General Counsel, Entergy Nuclear Operations R. Walker, Director, Radiation Control Program, Commonwealth of Massachusetts W. Irwin, Chief, CHP, Radiological Health, Vermont Department of Health The Honorable Therese Murray The Honorable Vincent deMacedo Chairman, Plymouth Board of Selectmen Chairman, Duxbury Board of Selectmen Chairman, Nuclear Matters Committee Plymouth Civil Defense Director
SUMMARY OF FINDINGS
.............................................................................................................. 3
REPORT DETAILS
.......................................................................................................................... 5
REACTOR SAFETY
........................................................................................................................ 5 1R01 Adverse Weather Protection ............................................................................... 5
1R04 Equipment Alignment .......................................................................................... 6
1R05 Fire Protection ..................................................................................................... 6
1R11 Licensed Operator Requalification Program ....................................................... 7 1R12 Maintenance Effectiveness ................................................................................. 8 1R13 Maintenance Risk Assessments and Emergent Work Control ........................... 9 1R15 Operability Evaluations ...................................................................................... 10 1R18 Plant Modifications ............................................................................................. 11 1R19 Post-Maintenance Testing
................................................................................ 12 1R20 Refueling and Other Outage Activities .............................................................. 12 1R22 Surveillance Testing .......................................................................................... 13
RADIATION SAFETY
.................................................................................................................... 13 2OS1 Access Control to Radiologically Significant Areas .......................................... 13 2OS2 As Low As Reasonably Achievable (ALARA) Planning and Controls .............. 14 2OS3 Radiation Monitoring Instrumentation and Protective Equipment ................... 14
OTHER ACTIVITIES
[OA] ............................................................................................................. 15
4OA1 Performance Indicator (PI) Verification ............................................................. 15 4OA2 Identification and Resolution of Problems ......................................................... 16 No findings of significance were identified. ....................................................................... 16 4OA3 Event Follow-up ................................................................................................. 20 4OA5 Other Activities ................................................................................................... 25 4OA6 Meetings, Including Exit ..................................................................................... 26
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
.......................................................................................... A-1
LIST OF ITEMS
OPENED, CLOSED AND DISCUSSED ............................................... A-1
LIST OF DOCUMENTS REVIEWED
.............................................................................. A-2
LIST OF ACRONYMS
..................................................................................................... A-9
Enclosure
- OF [[]]
- FINDIN [[]]
GS IR 05000293/2008-005; 10/01/2008-12/31/2008; Pilgrim Nuclear Power Station; Maintenance
Risk Assessments and Emergent Work Control, and Event Followup
The report covered a three month period of inspection by resident and region-based inspectors.
Two Green findings, both of which were non-cited violations (NCVs), were identified. The
significance of most findings is indicated by their color (Green, White, Yellow, Red) using
Inspection Manual Chapter (IMC) 0609, "Significance Determination Process" (SDP). Findings
for which the
NRC
management review. The NRC's program for overseeing the safe operation of nuclear power
reactors is described in
- NUR [[]]
EG-1649, "Reactor Oversight Process," Revision 4, dated
December 2006.
- A. [[]]
NRC-Identified and Self-Revealing Findings Cornerstone: Mitigating Systems
- Green. The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50.65(a)(4) for Entergy's failure to conduct a risk assessment for emergent maintenance
on the High Pressure Coolant Injection (HPCI) system injection valve. Specifically, the
failure to conduct a risk assessment resulted in Entergy not recognizing an increase in
risk to a Yellow condition, and therefore no risk management actions were taken.
Entergy entered this issue into their corrective action program. Corrective actions will
include revising attachments in Entergy's Technical Specification requirements
procedure to perform a risk review as a result of emergent maintenance activities. This finding was more than minor because Entergy failed to consider the unavailability of a risk significant system where the outcome of the risk assessment would have been a
change in a risk management category. The inspectors conducted an evaluation in
accordance with IMC 0609, "Significance Determination Process," Appendix K,
"Maintenance Risk Assessment and Risk Management Significance Determination
Process." The finding was determined to be of very low safety significance (Green)
because the Incremental Core Damage Probability Deficit for the timeframe that
- HP [[]]
CI was removed from service was significantly less than 1E-6. The inspectors determined
that this finding had a cross-cutting aspect in the area of Human Performance, Decision
Making, because Entergy did not use a systematic process to make a risk-significant
decision when faced with an unexpected plant condition. H.1(a) (Section 1R13) * Green. A self-revealing Green non-cited violation (NCV) of TS 5.4.1, "Procedures", was identified for a procedure which resulted in an inadvertent isolation of the Reactor Core
Isolation Cooling (RCIC) system. Specifically, the procedure was previously revised and
a step was inadvertently placed out-of-order. The procedure incorrectly instructed
technicians to remove relay contact blockers, or "boots", before clearing an isolation
signal which resulted in the system isolation. Entergy entered this issue into their
corrective action program. Corrective actions will include revising this procedure and reviewing other surveillance procedures that had been revised at the same time.
Enclosure This finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone. Isolating the
- RC [[]]
IC system
affected the cornerstone objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated this
finding using IMC 0609.04, "Phase 1 - Initial Screening and Characterization of
Findings". This finding was of very low safety significance because it was not a design or
qualification deficiency, did not represent a loss of system safety function, did not
represent an actual loss of a single train system for greater than the Technical
Specification allowed outage time, and was not made risk-significant because of external
events. The inspectors determined that this finding had a cross-cutting aspect in the
area of Human Performance, Resources, because Entergy did not ensure that the
procedure was complete and accurate. H.2(c) (Section 4OA3)
B. Licensee-Identified Violations None.
Enclosure
- REPORT [[]]
DETAILS Summary of Plant Status
Pilgrim Nuclear Power Station (PNPS) operated at or near 100 percent power during the majority
of the inspection period. However, on December 19, 2008, Entergy scrammed from 100 percent
power due to a load reject during a winter storm. Entergy resumed 100 percent power operation
on December 24, 2008. The plant remained at or near 100 percent for the remainder of the
inspection period.
1.
SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity 1R01 Adverse Weather Protection (71111.01)
.1 Seasonal Susceptibility a. Inspection Scope (1 sample)
The inspectors reviewed actions taken by the licensee in preparation for the onset of cold
weather during the week of November 2, 2008. The inspectors reviewed Procedure
8.C.40, Seasonal Weather Surveillance, and verified that selected steps had been completed. The inspectors walked down selected areas addressed in the procedure to
determine if heat tracing as well as plant heating systems were properly working. The
inspectors also walked down exterior portions of the Condensate Storage Tanks and the
Station Blackout Diesel. The documents reviewed during the inspection are listed in the
Attachment.
b. Findings
No findings of significance were identified.
.2 Impending Storm
a. Inspection Scope (1 sample) On December 19, 2008, a significant winter storm was tracking to impact the Pilgrim
plant that afternoon and into the evening. The inspectors reviewed Entergy's
preparations for the impending snow storm as well as for the high winds expected to
accompany the storm. The inspectors reviewed Entergy's severe weather procedures
including coastal storm preparations and operations during severe weather (specifically,
snow storm preparations). The inspectors also reviewed the stated plant risk given the
external risk increase and compared this to equipment that was out of service to
determine if there was an overall increase in risk. The inspectors conducted 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 or if there were any vulnerabilities to
snow accumulation (such as emergency diesel generator ventilation), and thereby impact
to safety related equipment. The documents reviewed during the inspection are listed in
Enclosure the Attachment.
b. Findings
No findings of significance were identified.
1R04 Equipment Alignment (71111.04)
Partial System Walkdowns (71111.04Q)
a. Inspection Scope (5 samples)
The inspectors performed five partial system walkdowns during this inspection period.
The inspectors reviewed the documents listed in the Attachment to determine the correct
system alignment. The inspectors conducted 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 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 were reviewed based on their risk significance for the
given plant configuration: * Instrument and Service Air Systems During Maintenance on K-111 Air Compressor; * Automatic Depressurization System with Reactor Core Isolation Cooling unavailable due to maintenance; * "B" Spent Fuel Pool System while in standby; * "A" train Salt Service Water (SSW) System with "D" SSW unavailable; and * "B" Core Spray System with High Pressure Coolant Injection out for maintenance.
b. Findings No findings of significance were identified.
1R05 Fire Protection (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
required 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, fire barriers, and any related compensatory measures. 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 the inspection are listed in the Attachment. The fire
Enclosure protection areas reviewed were: $ Reactor Building/El.23'-0" up to El.51'-0", East Side - Fire Area 1.9, Fire Zone 1.9; $ "A" Switchgear and Load Center Room - Fire Area 1.9, Fire Zone 2.2; $ "B" Switchgear and Load Center Room - Fire Area 1.10, Fire Zone 2.1; $ Spent Fuel Pool Cooling Pumps and Heat Exchanger Area - Fire Area 1.9, Fire Zone 1.13; and $ Vital Motor Generator Set Room - Fire Area 1.9, Fire Zone 3.5. b. Findings No findings of significance were identified.
1R11 Licensed Operator Requalification Program (71111.11)
.1 Licensee-Administered Annual Operating Tests
a. Inspection Scope (1 sample) On November 17, 2008, a region-based inspector conducted an in-office review of
results of the licensee-administered annual operating tests and comprehensive written
exams for 2008. The inspection assessed whether pass rates were consistent with the
guidance of NRC Manual Chapter 0609, Appendix I, "Operator Requalification Human
Performance Significance Determination Process." The inspector verified that: * Crew failure rate was less than 20 percent. (Crew failure rate was 0 percent) * Individual failure rate on the dynamic simulator test was less than or equal to 20 percent. (Individual failure rate was 0 percent) * Individual failure rate on the walk-through test was less than or equal to 20 percent. (Individual failure rate was 0 percent) * Individual failure rate on the comprehensive written exam was less than or equal to 20 percent. (Individual failure rate was 0 percent) * Overall pass rate among individuals for all portions of the exam was greater than or equal to 75 percent. (Overall pass rate was 100 percent)
b. Findings No findings of significance were identified.
.2 Licensed Operator Training
a. Inspection Scope (1 sample) The inspectors observed licensed operator training on November 18, 2008. Specifically,
the inspectors observed classroom Senior Reactor Operator (SRO) training on the
Enclosure Severe Accident Guidelines (SAGs), Core Mitigation Strategies, and the Emergency Operating Procedures (EOP). The lectures discussed
EOP entry conditions,
roles and responsibilities for the SROs, and the phenomenology of severe accidents. The inspectors assessed the training to determine if the training adequately prepared the
SROs to determine what actions to take in a severe accident situation and when to enter
the SAGs. The inspectors reviewed the lesson plans and applicable training objectives
to determine if they had been achieved. The documents reviewed during the inspection
are listed in the Attachment.
b. Findings No findings of significance were identified.
1R12 Maintenance Effectiveness (71111.12Q)
.1 Review of Functional Failures a. Inspection Scope (3 samples) The inspectors reviewed three functional failure determinations conducted in accordance
with Entergy procedures and the requirements of 10 CFR 50.65, Requirements for
Monitoring the Effectiveness of Maintenance at Nuclear Power Plants. The inspectors
reviewed the system maintenance rule functions, the basis for the conclusion that the
issues were considered functional failures, and the potential for common cause and
extent of condition. The inspectors also reviewed data to verify whether or not the
functional failures resulted in placing the systems in (a)(1). The inspectors reviewed
system health reports to determine if actions taken were reasonable and appropriate. In
addition, the inspectors reviewed Entergy's condition reports and corrective actions. The
documents reviewed during the inspection are listed in the Attachment. The functional
failure determinations reviewed were: *
PNP-2008-03338, High Pressure Coolant Injection Valve relay in circuit breaker cabinet fails.
b. Findings No findings of significance were identified.
.2 Review of K-117 Air Compressor (a)(1) Action Plan a. Inspection Scope (1 sample) The inspectors reviewed the (a)(1) corrective action plan for the K-117 air compressor
unavailability exceeding the (a)(2) unavailability criteria for items such as: (1) appropriate work practices; (2) identifying and addressing common cause failures; (3) scoping in
accordance with
MR); (4) characterizing
reliability issues for performance; (5) trending key parameters for condition monitoring; (6) charging unavailability for performance; (7) classification and reclassification in
Enclosure accordance with
- 10 CFR 50.65(a)(1) or (a)(2); and (8) appropriateness of performance criteria for structures, systems, and components (
SSCs)/functions classified as (a)(2)
and/or appropriateness and adequacy of goals and corrective actions for SSCs/functions
classified as (a)(1). The documents reviewed during the inspection are listed in the
Attachment.
b. Findings No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13)
a. Inspection Scope (3 samples) The inspectors evaluated three online 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 verified the appropriate use of Entergy's
risk assessment tool, Equipment Out of Service (EOOS), and entry into appropriate risk
categories. 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 documents reviewed during the
inspection are listed in the Attachment. The inspectors reviewed the conduct and
adequacy of maintenance risk assessments for the following maintenance and testing
activities: $ Yellow Risk, During Reactor Core Isolation Cooling and Air Compressor K-111 Maintenance and Testing Activities; $ Emergent Risk of Inoperability of the High Pressure Coolant Injection and Reactor Core Isolation Cooling Systems; and $ Forced Outage Shutdown Risk Assessments.
b. Findings Introduction. The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50.65(a)(4) for Entergy's failure to conduct a risk assessment for emergent maintenance
on the High Pressure Coolant Injection (HPCI) system injection valve. Specifically, the
failure to conduct a risk assessment resulted in Entergy not recognizing an increase in
risk to a Yellow condition, and therefore no risk management actions were taken.
Description. At 7:44 p.m. on October 21, 2008, operators received an alarm in the control room and determined the cause was a loss of control power for a
- HP [[]]
CI injection
valve. Entergy declared
- HP [[]]
CI inoperable and entered the Limiting Condition for
Operation (LCO) for Technical Specification (TS) 3.5.C.2, "HPCI System." This LCO
requires that the system be made operable within 14 days, otherwise be in cold
shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
With
- HP [[]]
CI unavailable, risk management actions are generally taken to protect those
systems that provide redundancy for its function, such as the Reactor Core Isolation
Enclosure Cooling system. Entergy did not conduct a risk assessment nor recognize the plant risk condition was "Yellow" and therefore did not take any risk management actions. Entergy
restored
LCO eight hours later.
Analysis. The performance deficiency associated with this finding is that Entergy did not perform a review of the increased risk while
- HP [[]]
CI was inoperable and, as a result, did
not take risk management actions as required by 10 CFR 50.65(a)(4). This finding is
associated with the human performance attribute of the Mitigating Systems cornerstone
and is more than minor because Entergy failed to consider the unavailability of a risk significant system where the outcome of the risk assessment would have been a change
in risk management category. The inspectors conducted an evaluation in accordance
with IMC 0609, "Significance Determination Process," Appendix K, "Maintenance Risk
Assessment and Risk Management Significance Determination Process." The finding
was determined to be of very low safety significance (Green) because the Incremental
Core Damage Probability Deficit for the timeframe that
- HP [[]]
CI was removed from service was significantly less than 1E-6.
The inspectors determined that this finding had a cross-cutting aspect in the area of
Human Performance, Decision Making, because Entergy did not use a systematic
process to make a risk-significant decision when faced with an unexpected plant
condition. H.1(a)
Enforcement. 10 CFR 50.65(a)(4), "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," states, in part, that "...the licensee shall assess and manage the increase in risk that may result from the proposed maintenance
activities." Contrary to the above, from October 21, 2008 to October 22, 2008, Entergy
failed to assess the increased risk that resulted from
- HP [[]]
CI unavailability. As a result, Entergy did not recognize a "Yellow" risk condition and did not take any risk management
actions. Corrective actions will include revising attachments in Entergy's Technical Specification requirements procedure to perform a risk review as a result of emergent
maintenance activities. Because this violation was of very low safety significance
(Green) and was entered into the licensee's corrective action program (CR-PNP-2008-
03792), this violation is being treated as an
NCV 05000293/2008005-01, Failure to Conduct a Risk Assessment for Emergent Maintenance on the High Pressure Coolant Injection
System) 1R15 Operability Evaluations (71111.15)
a. Inspection Scope (5 samples) The inspectors reviewed five operability determinations associated with degraded or non-conforming conditions to determine if the operability determination was justified and if the mitigating systems or those affecting barrier integrity remained available 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 licensee performance against
related Technical Specification and
- UFS [[]]
AR requirements. The documents reviewed
during the inspection are listed in the Attachment. The inspectors reviewed the following
degraded or non-conforming conditions:
Enclosure $
- CR -PNP-2008-03404, Shutdown Transformer Breaker (A802) won't close remotely from the Control Room; $
PNP-2008-03611, Thermography identifies a hot spot in Air Cooled Breaker (ACB) 104.
b. Findings No findings of significance were identified.
1R18 Plant Modifications (71111.18)
.1 Permanent Modification a. Inspection Scope (1 sample) The inspectors reviewed Permanent Modification
CFR 50.59 screening, to determine whether the
licensing bases and performance capability of the associated system had been degraded through the modification. A walkdown of the "A" Residual Heat Removal Heat Exchanger
was performed to determine if the PSV-8008 valve's new orientation would be subject to
additional stress or be impacted by other adverse conditions. The inspectors reviewed
system drawings to determine whether they reflected the permanent modification. The
documents reviewed during the inspection are listed in the Attachment.
b. Findings No findings of significance were identified.
.2 Temporary Modification a. Inspection Scope (1 sample) The inspectors reviewed Temporary Modification Engineering Change (EC) 7768,
"Temp. Mod. Required to Document Heat Detection System Disabled on Pre-Action
Sprinkler System for the Turbine Bearings", to determine whether the performance
capability of the Fire Protection System had been degraded through the modification. The inspectors reviewed the Updated Fire Hazards Analysis, procedures, and the 10
CFR 50.59 screening to ensure the temporary modification did not adversely affect fire
protection program attributes. The inspectors reviewed control room drawings to
determine whether they properly reflected the temporary modification. The documents
reviewed during the inspection are listed in the Attachment.
b. Findings No findings of significance were identified.
Enclosure 1R19 Post-Maintenance Testing (71111.19)
a. Inspection Scope (6 samples) The inspectors reviewed six 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 TS
requirements. The inspectors also evaluated whether conditions adverse to quality were
entered into the corrective action program for resolution. The documents reviewed
during the inspection are listed in the Attachment. The following maintenance activities
and their post-maintenance tests were evaluated: $ Reactor Core Isolation Cooling Outboard Isolation Valve Electrical Maintenance; $ Open and Inspect Residual Heat Removal (RHR) Pump Discharge Check Valve 1001-67A; $ Replace Undervoltage Relay for High Pressure Coolant Injection Pump Injection Valve
RHR Pump "A" Relays; $ Overhaul of Salt Service Water Pump 208D; and $ Replace Current Transformers on the F-15 Substation Feeding the Shutdown Transformer.
b. Findings No findings of significance were identified.
1R20 Refueling and Other Outage Activities (71111.20)
a. Inspection Scope (1 sample) The inspectors reviewed the outage plan and shutdown risk assessments for a forced,
non-refueling outage conducted from December 19, 2008, through December 23, 2008.
The outage was conducted following a plant transient due to a load reject and
subsequent reactor plant scram. The load reject was the result of a significant fault in a
switchyard breaker during a severe winter storm. During this outage, the inspectors
observed plant shutdown activities including the outage activities listed below. The
documents reviewed during the inspection are listed in the Attachment.
- Hot Shutdown Control * Shutdown Risk Assessment and Risk Management * Implementation of TS * Outage Control Center Activities * Plant Startup * Licensee identification and resolution of problems identified during and related to outage activities
Enclosure b. Findings No findings of significance were identified.
1R22 Surveillance Testing (71111.22)
a. Inspection Scope (2 samples) The inspectors reviewed two samples of surveillance activities 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 accordance with the procedural steps. Additionally, the inspectors
evaluated the applicable test acceptance criteria for consistency with associated design
bases, licensing bases, and TS requirements. The inspectors also evaluated whether
conditions adverse to quality were entered into the corrective action program for
resolution. The documents reviewed during the inspection are listed in the Attachment.
The following surveillance tests were evaluated: $ Standby Liquid Control In-Service Testing (IST); and $ "B" Emergency Diesel Generator Operability Testing. b. Findings No findings of significance were identified. 2.
- RADIAT [[]]
- ION [[]]
OS1 Access Control to Radiologically Significant Areas (71121.01) a. Inspection Scope (1 sample) During the period of October 20 through 23, 2008, the inspectors conducted the following
activities to verify that the licensee was properly implementing physical, administrative,
and engineering controls for access to locked high radiation areas, and other
radiologically controlled areas (RCA) during power operations. Implementation of these
controls was reviewed against the criteria contained in 10 CFR 20, relevant Technical
Specifications, and the licensee=s procedures. This inspection activity represents the completion of one (1) sample relative to this inspection area. Plant Walkdown and Radiation Work Permits (RWP) Reviews The inspectors examined Pilgrim's physical and programmatic controls for highly
activated or contaminated materials (non-fuel) stored within the spent fuel pool. The
inspectors toured the spent fuel pool area and reviewed the procedure for handling
highly radioactive objects. The inspectors also toured the Traversing Incore Probe (TIP)
Room and Torus Room areas. The inspectors observed the postings and barricades in
each area and reviewed surveys and RWPs for the areas including the electronic
Enclosure personal dosimeter alarm set points (both integrated dose and dose rate) for conformity with survey indications and plant policy.
b. Findings No findings of significance were identified.
ALARA) Planning and Controls (71121.02) a. Inspection Scope (5 samples) During the period October 20 through 23, 2008, the inspector conducted the following
activities to verify that the licensee was properly implementing operational, engineering,
and administrative controls to maintain personnel exposure
- ALA [[]]
RA during routine plant
operation. Implementation of these controls was reviewed against the criteria contained
in 10 CFR 20, applicable industry standards, and the licensee=s procedures. This represents the completion of five samples relative to this inspection area.
Inspection Planning The inspectors requested a list of the work activities ranked by actual exposure that were
completed during refueling outage (RFO) 16. The inspectors reviewed the
RFO16. The inspectors
reviewed the
- ALA [[]]
RA work activity evaluations, exposure estimates, and exposure
mitigation requirements. The inspectors reviewed the exposure estimates and compared
the estimates with the actual dose received.
Radiation Worker Performance The inspectors observed radiation worker performance prior to entering the RCA. The
inspectors questioned workers relative to their individual and department dose goals and
their understanding of the previous day's goals and the actual dose received.
Declared Pregnant Workers
The inspectors requested exposure results and monitoring controls employed for
declared pregnant workers with respect to the requirements of 10 CFR 20. There were
no declared pregnant workers since January 1, 2008.
b. Findings No findings of significance were identified. 2OS3 Radiation Monitoring Instrumentation and Protective Equipment (71121.03) a. Inspection Scope (1 sample) During the period of October 20 through 23, 2008, the inspectors conducted the following
activities to evaluate the adequacy of the licensee's program to maintain Self Contained
Breathing Apparatus (SCBA). This inspection activity represents the completion of one
15sample relative to this inspection area.
Self-Contained Breathing Apparatus (SCBA) Maintenance and User Training
The inspectors reviewed the status and surveillance records of
SCBA in the control room
and compared the records with the actual equipment staged for use. The inspectors
verified that control room operators and other emergency response personnel are
trained.
b. Findings No findings of significance were identified. 4.
- OTHER [[]]
- ACTIVI [[TIES [OA]]]
PI) Verification (71151)
.1 Mitigating Systems a. Inspection Scope (2 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, CRs, System Health Reports,
and NRC inspection reports. The acceptance criteria used for the review was Nuclear
Energy Institute (NEI) 99-02, Revision 5, "Regulatory Assessment Performance Indicator
Guidelines." Documents reviewed during the inspection are listed in the Attachment.
The following performance indicators were reviewed: * Emergency Diesel Generators (EDG) from the fourth quarter 2007, through the third quarter of 2008; and * Cooling Water (Salt Service Water/RBCCW) from the fourth quarter 2007, through the third quarter of 2008.
b. Findings No findings of significance were identified.
.2 Occupational Exposure Control Effectiveness
a. Inspection Scope (1 sample) The inspectors reviewed implementation of the licensee=s Occupational Exposure Control Effectiveness Performance Indicator (PI) Program. Specifically, the inspector reviewed recent
- ACTI [[]]
ON reports, and associated documents, for occurrences involving
locked high radiation areas, very high radiation areas, and unplanned exposures against
the criteria specified in Nuclear Energy Institute (NEI) 99-02, Revision 5, "Regulatory
16Assessment Performance Indicator Guidelines," to verify that all occurrences that met the NEI criteria were identified and reported as performance indicators. This inspection
activity represents the completion of one sample relative to this inspection area;
completing the annual inspection requirement.
b. Findings No findings of significance were identified.
.3
ODCM Radiological Effluent Occurrences
a. Inspection Scope (1 sample) The inspectors reviewed relevant effluent release reports for the period of January 1,
2007, through December 31, 2007, for issues related to the public radiation safety
performance indicator, which measures radiological effluent release occurrences that
exceed 1.5 millirem / quarter whole body or 5.0 millirem / quarter organ dose for liquid effluents; 5 millirads / quarter gamma air dose, 10 millirads / quarter beta air dose, and 7.5 millirads / quarter for organ dose for gaseous effluents. This inspection activity represents the completion of one sample relative to this inspection area; completing the
annual inspection requirement.
b. Findings No findings of significance were identified.
- 4OA 2 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 the licensee's CAP.
This review was accomplished by reviewing printouts of each CR, attending daily
screening meetings and/or accessing the licensee's database. The purpose of this
review was to identify conditions such as repetitive equipment failures or human
performance issues that might warrant additional follow-up.
b. Findings No findings of significance were identified.
.2 Annual Sample: Operator Workarounds
a. Inspection Scope (1 sample) The inspectors performed the annual review of operator workarounds to verify Entergy
17was identifying operator workaround problems at an appropriate threshold and entering them into the corrective action program. The inspectors reviewed identified workarounds
to determine whether the mitigating system function was affected, whether the operator's
ability to implement abnormal and emergency operating procedures was affected, and whether appropriate procedures had been updated to reflect actual plant conditions. The
inspection was accomplished through personnel interviews, plant tours, and review of
station documents. The documents reviewed during the inspection are listed in the
Attachment.
b. Findings and Observations No findings of significance were identified. Operator workarounds have been identified
and entered into the corrective action program for resolution. No unrecognized impacts
to operator or system performance were identified, and corrective actions have been
implemented to restore the affected systems.
.3 Annual Sample: Safety Relief Valve Leakage a. Inspection Scope (1 sample) The inspectors selected the issue of Safety Relief Valve (SRV) leakage as an inspection
sample for in-depth review because of recent forced plant shutdowns due to SRV
leakage. Additionally, SRV leakage has been a long-standing issue at Pilgrim Nuclear
Power Station (PNPS). This inspection was conducted to determine if Entergy was
taking appropriate corrective actions to address SRV leakage.
The inspectors reviewed procedures, condition reports, engineering evaluations, root
cause analyses, and interviewed plant personnel to assess Entergy's problem
identification, evaluation, and corrective action effectiveness with respect to SRV
leakage. Additionally, the inspectors reviewed the Technical Specifications and Updated
Final Safety Analysis Report to assess the adverse impact of SRV leakage with respect
to design basis requirements. The documents reviewed during this inspection are listed
in the Attachment. b. Findings and Observations No findings of significance were identified.
Pilgrim Nuclear Power Station (PNPS) Safety Relief Valves are of the two-stage Target
Rock-type design, consisting of a pilot-stage assembly and a main-stage assembly.
Industry Operating Experience has shown that two-stage Target Rock SRVs exhibit
some amount of pilot-stage leakage during plant operation. Additionally, industry
operating experience has quantified
SRV tailpipe temperature,
as well as an upward setpoint drift impact.
- PN [[]]
PS Technical Specifications require that
an engineering evaluation be performed to justify continued operation with elevated
tailpipe temperatures. On a number of such occasions, Entergy has implemented an
Operational Decision Making Issue (ODMI) to establish administrative limits for the
maximum allowable SRV tailpipe temperature for continued plant operation. The
18operating limits were established to maintain
PNPS throughout the plant's operating history.
Entergy has been forced to shutdown the plant from full power operation on three
occasions. On each occasion, SRV tailpipe temperatures approached the administrative
limits imposed in the respective
- OD [[]]
MIs for March 2004, December 2007, and April 2008.
Following the March 2004 shutdown, the leaking valves were removed from service and
a root cause analysis was performed as a result of high, as-found setpoint testing results.
The root cause was determined to be corrosion bonding of the pilot valve disc/seat, with
a contributing cause of insulation deficiencies. The contributing cause determination
stated that proper fitting
SRVs, to reduce the
propensity of leakage and eliminate the conditions conducive to corrosion binding.
Corrective actions included refurbishing the pilot valve discs, as well as SRV insulation
enhancements. The December 2007 shutdown also identified inadequate SRV fitting
insulation as one of the root causes for the
CR-PNP-2007-04936). The
inspector noted Entergy has self-identified deficiencies with SRV insulation dating back
to 1993.
The December 2007, and April 2008, shutdowns identified inadequate simmer margin as
one of the root causes (CR-PNP-2007-04936) for SRV pilot-stage leakage. Simmer
margin is defined as the pressure difference between SRV setpoint and plant normal
operating pressure. At
- PN [[]]
PS, the plant simmer margin is limited by a +/-1% Technical
Specification tolerance (i.e., margin) for the allowable
accident analyses have shown this tolerance is required to maintain peak reactor vessel
pressures within the code allowable limits. The inspector noted Entergy has self-
identified insufficient simmer margin in a number of condition reports dating back to
2004. Additionally, General Electric has issued Service Information Letters to highlight
the relationship between pilot-stage leakage and plant operating simmer margin. At the
time of this inspection, Pilgrim was investigating corrective action options to address
simmer margin, as documented in
PNP-2007-04936. The options being pursued
included contracting with vendors to increase plant simmer margin via analysis and/or
modifications, and completing an independent root cause analysis of pilot valve disc/seat
leakage. Entergy was also pursuing the option for development of a new pilot valve
disc/seat design.
At
SRV upward setpoint drift, even due to small amounts of
leakage, are magnified by the limited
TS due to the
designed plant relief capacity. The inspector found Entergy's planned corrective actions,
to address the contributing elements of SRV leakage, to be appropriate.
19.4 Annual Sample: Review of Aggregate Impact of Significant Events Which Had Occurred During 2007 a. Inspection Scope (1 sample) The inspectors selected
PNP-2007-04865 for detailed review. The CR was written in December 2007, to evaluate whether there were any additional insights or trends to be
identified from an aggregate review of significant events which had occurred during 2007.
The inspectors reviewed the licensee's analysis and recommendations for corrective
actions. The documents reviewed during the inspection are listed in the Attachment. b. Findings and Observations No findings of significance were identified. Entergy reviewed significant events which
occurred during 2007, and categorized the events as to their root and contributing causes,
whether any regulatory findings had been issued, their impact on plant operation, and the
aggregate impact on plant performance (i.e., equipment or human performance). Entergy
determined that several of the issues related to the performance of the Emergency Diesel
Generators (EDG) and conducted a separate analysis of these events. TI-176 has been
conducted to evaluate current EDG testing, the results of which are documented in Section
4OA5. Entergy also reviewed the remaining events and determined that, in some cases,
expanding the scope of the evaluation of a given event or condition to review other systems
or programs would have been appropriate. As a result, Entergy has instituted a corrective
action for the Operations Department to include, as part of their review of degraded
condition operability, the need to consider other systems or components that may have a similar vulnerability to the degraded condition. In addition, Entergy will be conducting training with Engineering Department staff to similarly consider other systems, programs, or
components when evaluating degraded conditions for extent of condition and corrective
actions.
The inspectors reviewed Entergy's corrective actions and evaluated more recent issues to
determine if Entergy applied the principles discussed above. The inspectors noted
instances where both Operations Department and Engineering Department staffs evaluated
a given issue for its applicability beyond the condition itself and, as a result, specified additional corrective actions. The inspectors concluded that these initiatives should improve
Entergy's effectiveness in evaluating and correcting issues and in identifying other areas for
improvement that may not have been previously considered.
.5 Semi-Annual Review to Identify Trends a. Inspection Scope (1 sample) The inspectors performed a review of Entergy's Corrective Action Program (CAP) and
associated documents to identify trends that could indicate the existence of a more
significant safety issue. The review was focused on repetitive equipment and corrective
maintenance issues, but also considered the results of daily inspector CAP item screening.
The review included issues documented in
CAP performance
indicator data. The review focused on the six month period of July 2008, through December
202008, although the inspectors also evaluated previous trend results for
NRC Inspection Reports 05000293/2008003
and 05000293/2007005. Documents reviewed during the inspection are listed in the
attachment.
b. Findings and Observations No findings of significance were identified. In NRC Inspection Report 05000293/2008003,
the inspectors concluded that corrective actions to improve configuration control at Pilgrim
had not been in effect long enough to conclude whether they were effective. As a result, the
configuration control low level trend originally initiated in the fourth quarter report of 2007,
continued to be monitored during the past two quarters. The inspectors noted that Entergy's
corrective actions appear to have been effective in reducing the number of mispositioning
errors. Corrective actions have included conducting operator fundamental training on
precise plant control and human performance, generating lessons learned, and generating a
standard for mispositioned components and a fleet procedure on configuration control.
In addition to the corrective actions discussed in NRC Inspection Report 05000293/2008003,
the Operations department instituted a Mentorship Pilot Program to augment additional
reactor operators to assist non-licensed operators in their professional development.
Entergy has focused their program in the areas of operator fundamentals, human
performance, shift turnover, preparation for shift activities, and operator engagement during
briefings. Corrective actions undertaken by the Operations Department appear to have
improved the performance of configuration control and, as a result, this low level trend is
considered closed. No additional low level trends were identified which would indicate the
presence of a broader safety issue.
4OA3 Event Follow-up (71153)
.1 Unplanned Reactor Core Isolation Cooling (RCIC) Isolation
a. Inspection Scope (1 sample) On the evening of October 6, 2008, operators inadvertently isolated the
- RC [[]]
IC system
while performing Surveillance Procedure
RCIC Steam Line High
Temperature Instrument Functional Test. Specifically, "boots" which were installed to
block applicable relays were removed before auto-isolation trip signals had cleared. The
- RC [[]]
IC system isolated due to a Group 5 signal which closed the inboard and outboard
steam supply valves, which resulted in the actuation of the turbine trip throttle valve. The
Group 5 isolation was reset and the operators placed the
- RC [[]]
IC system back in its
standby lineup.
The inspectors reviewed operator logs, applicable procedural requirements, and
technical specifications. The documents reviewed during the inspection are listed in the
Attachment.
21 b. Findings See Section 4OA3.5.
.2 Entergy Response to a Fire in the Health Physics Calibration Laboratory
a. Inspection Scope (1 sample) On October 29, 2008, a fire occurred in the Health Physics (HP) instrumentation
calibration laboratory located in the Operations and Maintenance (O&M) building.
Following the start of the electric fire pump and the generation of O&M building smoke
alarms, the on-site Fire Brigade identified that a fire had occurred in the HP calibration
lab and that the fire suppression system had actuated and extinguished the fire.
Operators requested off-site assistance from the Plymouth Fire Department and both
organizations entered the room to verify that the fire suppression system had
extinguished the fire. In addition, Entergy declared an Unusual Event due to the
occurrence of a fire in the protected area for which off-site assistance was requested.
The inspectors responded to the site to evaluate Entergy's actions in response to the fire
and to assess any impact on the licensed radiological materials located in the HP
calibration laboratory. The inspectors reviewed Entergy's immediate actions to respond
to the fire, root cause investigation, event timeline, and corrective actions. The
documents reviewed during the inspection are listed in the Attachment.
b. Findings No findings of significance were identified.
.3 Operator Response to Unplanned Reactor Core Isolation Cooling (RCIC) Unavailability
a. Inspection Scope (1 sample) On October 22, 2008, operators determined that the
- RC [[]]
IC flow controller was inoperable
due to aged power supply capacitors. The industry recommended service life is 7 to 10
years and the
- RC [[]]
IC capacitors had been in service from 21 to 30 years. Operations
declared
TS 3.5.D, "Reactor Core Isolation Cooling (RCIC)
System", a 14-day shutdown action statement. The control panel flow controller was
replaced with the flow controller from the alternate shutdown panel (ASP) and operations
entered
ASP. During post-installation
testing, the flow controller from the ASP was not able to maintain rated flow at the
required pressure. A refurbished controller intended to replace the one from the ASP
was then installed in the control panel. This flow controller was able to achieve and
maintain rated pressure and flow. Operations then declared the system operable and
exited
CFR 50.72, 8-hour, non-emergency notification
report for a condition that could have prevented the fulfillment of the safety function of a mitigating system. Entergy conducted subsequent bench testing of the aged capacitors
from November 6, 2008 to November 8, 2008 and determined that these capacitors
would have been able to perform their function in the required mission time. As a result,
Entergy retracted their 10 CFR 50.72 report on December 9, 2008. The inspectors
2reviewed Technical Specifications, control room logs, interviewed operations and engineering personnel, and reviewed the basis for the retraction of the 10 CFR 50.72
report. The documents reviewed during the inspection are listed in the Attachment.
b. Findings No findings of significance were identified.
.4 Unplanned High Pressure Coolant Injection (HPCI) Isolation a. Inspection Scope (1 sample) On the afternoon of November 20, 2008, operators inadvertently isolated the
- HP [[]]
HPCI High
Steam Line Temperature. Specifically, in-series, high temperature relays were both
inadvertently opened causing a Group 4 isolation signal. A Group 4 isolation signal
closes both the inboard and outboard steam supply valves, which rendered the
- HP [[]]
- HP [[]]
CI
system back in its standby lineup using Procedure 2.2.21, High Pressure Coolant
Injection. The inspectors reviewed operator logs, applicable procedural requirements,
and Technical Specifications. The documents reviewed during the inspection are listed
in the Attachment.
b. Findings
No findings of significance were identified.
.5 Reactor Scram on Load Reject from failure of 345 KV Switchyard Circuit Breakers 104 and 105 a. Inspection Scope (1 sample) On December 19, 2008, operators responded to a reactor scram from 100% reactor
power due to a load reject from the main generator. During a coastal winter storm, an
electrical fault developed on the transformer side of circuit breakers 104 and 105
(flashover of the ACB-105 "A" phase generator bushing resulted in a significant current
to ground fault) resulting in the main transformer differential relay, main transformer
overcurrent relay, and main transformer distance relay actuations to open breakers 104
and 105. Without circuit breakers 104 and 105 to route power to the electrical grid, the
turbine, and hence, the reactor scrammed due to a main turbine trip initiated by a main
generator load reject. As a result of the reactor scram and load reject, three of four
safety relief valves opened briefly, which is expected for the condition. Entergy also
received a Group Two isolation signal for Secondary Containment and a Group Six
isolation signal for Reactor Water Cleanup. Both of these isolation signals were
expected for the situation. However, Y3 and Y4, two of Pilgrim's 120 VAC safety related
instrument buses, remained de-energized following the trip, an unexpected response.
Y3 and Y4 are 120 VAC electrical buses which power mitigating system instrumentation
including containment isolation logic, SRV acoustic monitoring, and other safety related
23equipment. This issue was investigated and corrected prior to startup. In addition, a 10 CFR 50.72 notification was generated due to the valid actuation of the Reactor
Protection System. The inspectors responded to the control room, reviewed reactor
plant parameters and operator response to this event. The documents reviewed during
the inspection are listed in the Attachment.
b. Findings No findings of significance were identified.
.6 Loss of Start-Up Transformer Power Supply from
ACB 102 and 103 a. Inspection Scope (1 sample) On December 20, 2008, operators responded to a momentary loss of the 345 KV power
supply from the ACB 102 and 103 switchyard circuit breakers. This resulted in the loss
of buses A1 through A6 and their associated loads. The emergency diesel generators
responded as designed to provide power to vital buses A5 and A6. Group Isolations 1, 2
and 6 were generated on the Reactor Protection Systems signal (Main Steam Isolation
Valves (MSIV) closed, primary sample valves isolated, Reactor Water Clean Up isolated
and the Reactor Building Ventilation system isolated). The Reactor Core Isolation
Cooling and High Pressure Coolant Injection Systems were operated in manual to
maintain reactor vessel level and pressure. ACB 102 and 103 re-closed and Buses A1
through A4 were recovered and Buses A5 and A6 were restored to the Start-Up
Transformer power supply.
- MS [[]]
IV's were opened and Group Isolation Signals were reset.
The licensee generated a 10 CFR 50.72 8-hour report to the Nuclear Regulatory
Commission discussing this event. The inspectors responded to the control room,
reviewed reactor plant parameters and operator response to this event. The documents
reviewed during the inspection are listed in the Attachment.
b. Findings
No findings of significance were identified.
.7 (Closed)
RCIC) System Declared Inoperable During Surveillance Testing due to Procedure Error
a. Inspection Scope (1 sample)
The inspectors evaluated
RCIC System Declared
Inoperable during Surveillance Testing Due to Procedure Error". This LER is closed with a finding.
b. Findings Introduction. A self-revealing Green non-cited violation (NCV) of TS 5.4.1, "Procedures", was identified for a procedure error which resulted in an inadvertent isolation of the
24Reactor Core Isolation Cooling (RCIC) system. Specifically, the procedure was previously revised and a step was inadvertently placed out-of-order which resulted in the
isolation of
- RC [[]]
RCIC Steam Line High
Temperature Instrument Functional Test". When performing this surveillance, Instrumentation and Controls (I&C) technicians use "boots" to block relay contacts and
prevent closure of the
- RC [[]]
IC steam supply line isolation valves. The I&C technicians
followed the procedure, which incorrectly instructed them to remove the "boots" before
clearing the auto-isolation trip signals that would close the valves. This resulted in a
Group 5 isolation signal that isolated the
- RC [[]]
IC system by closing the inboard and
outboard steam supply valves, actuating the
- RC [[]]
IC turbine trip throttle valve, and
rendering the
- RC [[]]
IC system inoperable. The operators reset the Group 5 isolation and
placed the
- RC [[]]
IC system back in its standby lineup within one hour.
The licensee had previously revised several surveillance procedures for the
HPCI) System. During the revision to
Procedure 8.M.2-2.6.3, the step to remove the "boots" was placed before the step to
clear the auto-isolation signal. Corrective actions will include revising this procedure and reviewing other procedures that had been revised at the same time.
Analysis. The performance deficiency associated with this finding was that Entergy introduced an error into a procedure during revision that resulted in an inadvertent
isolation of the
- RC [[]]
IC system. This finding was more than minor because it was
associated with the equipment performance attribute of the Mitigating Systems
cornerstone. Isolating the
- RC [[]]
IC system affected the cornerstone objective to ensure the
availability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated this finding using IMC 0609.04, "Phase 1 -
Initial Screening and Characterization of Findings". This finding was of very low safety
significance because it was not a design or qualification deficiency, did not represent a
loss of system safety function, did not represent an actual loss of a single train system for
greater than the TS allowed outage time, and was not made risk-significant because of
external events.
The inspectors determined that this finding had a cross-cutting aspect in the area of
Human Performance, Resources, because Entergy did not ensure that the procedure
was complete and accurate. H.2(c)
Enforcement. Technical Specification 5.4.1, "Procedures", requires that written procedures be maintained as recommended in
RG) 1.33,
"Quality Assurance Program Requirements," Revision 2, Appendix A, February 1978.
RG 1.33, Appendix A, Section 8 includes procedures for surveillance tests. Contrary to this, Procedure 8.M.2-2.6.3 was not adequately maintained, because it included an error
that resulted in an isolation of the
- RC [[]]
IC system. Corrective actions will include revising this procedure and reviewing other surveillance procedures that had been revised at the same time. Because this finding is of very low safety significance and Entergy has
entered it into their corrective action program (CR-PNP-2008-03182), this violation is
25being treated as an
- RC [[]]
- TS 4.0.3 When It Was Discovered That a Surveillance Had Never Been Performed On June 25, 2007, Entergy informed the Nuclear Regulatory Commission (
NRC) staff
that it had missed a Technical Specification (TS) surveillance requirement to perform time response testing of four Reactor Protection System (RPS) scram contactors. During
their review, Entergy identified that the four RPS scram contactors had never been
tested. Entergy evaluated the operability of the
TS 4.0.3, "Surveillance Requirement Applicability," would allow a delay period up to the limit of the specified surveillance frequency. The
inspectors questioned Entergy regarding the applicability of
RPS scram contactors, as compared to
missing a surveillance test following satisfactory initial system baseline testing that
originally showed system operability. As a result of this implementation of
TS 3.1, "Reactor Protective System,"
which constituted a violation of NRC requirements. Entergy later modified the applicable
surveillance procedures and successfully response time tested all RPS scram
In Task Interface Agreement (TIA) 2008-004, the NRC staff disagreed with Entergy on its
implementation of
"Reactor Protection System," as a result. Discretion is warranted because: (1) licensee
current basis documents do not specifically clarify the distinction between a missed
surveillance and one that has never been performed, (2) the licensee subsequently
completed the surveillance testing satisfactorily, and (3) the issue was of very low safety
significance, since when the correct testing was accomplished, it was completed
satisfactorily indicating that the timing of the reactor scram function was not negatively
impacted. Accordingly, the
violation in accordance with Section
- NRC Enforcement Policy and no violation will be issued. Enforcement Action (EA)09-013, Failure to Enter
URI 05000293/2007003-04 is closed.
2. Implementation of Temporary Instruction (TI) 2515/176 - Emergency Diesel Generator Technical Specification Surveillance Requirements Regarding Endurance and Margin Testing
a. Inspection Scope The objective of TI 2515/176, "Emergency Diesel Generator Technical Specification
Surveillance Requirements Regarding Endurance and Margin Testing," was to gather
information to assess the adequacy of nuclear power plant emergency diesel generator
26(EDG) endurance and margin testing as prescribed in plant-specific technical specifications (TS). The inspectors reviewed emergency diesel generator ratings, design
basis event load calculations, surveillance testing requirements, and emergency diesel generator vendor's specifications and gathered information in accordance with TI
2515/176. The inspector assessment and information gathered while completing this TI
was discussed with licensee personnel. This information was forwarded on to the Office
of Nuclear Reactor Regulation for further review and evaluation.
b. Findings No findings of significance were identified. 4OA6 Meetings, Including Exit On September 11, 2008, the inspectors presented the preliminary inspection results of a
Problem Identification and Resolution sample to Mr. B. Sullivan, Nuclear Engineering Director, Mr. S. Bethay, Safety Assessment Director, and other members of the Entergy
staff. Following additional in-office review, the inspectors conducted a final exit meeting
via teleconference on December 16, 2008, with Mr. S. Bethay and other members of
Entergy staff. The inspectors verified that no proprietary information is documented in
this report.
On October 23, 2008 at 9:30 A.M., the Occupational Radiation Safety exit meeting was
held by phone with Mr. Joe Lynch, Licensing Manager.
On October 31, 2008, an exit meeting of the results of Temporary Instruction (TI)
2515/176 was conducted. The preliminary inspection results were presented to
Mr. Stan Wollman, Engineering Supervisor, and other members of the Pilgrim staff. The
inspector confirmed that no proprietary information was provided or examined during the
inspection. On January 7, 2009, the resident inspectors conducted an exit meeting and presented
the preliminary inspection results to Mr. Kevin Bronson, Site Vice President, and other
members of the Pilgrim staff. The inspectors confirmed that no proprietary information
was provided or examined during the inspection.
- ATTACH [[]]
- MENT [[:]]
- SUPPLE [[]]
- MENTAL [[]]
- INFORM [[]]
- SUPPLE [[]]
- MENTAL [[]]
- INFORM [[]]
- ATION [[]]
- KEY [[]]
- POINTS [[]]
- OF [[]]
CONTACT Licensee personnel:
K. Bronson Site Vice President
R. Smith General Manager Pilgrim Operations
ALARA Technician
J. Lamoureux Sr. Project Manager
W. Lobo Licensing Engineer
J. Lynch Licensing Manager
W. Mauro Supervisor, Radiological Engineering
C. Minott Sr. Project Manager
D. Noyes Operations Manager
R. O'Neill Operations Outage Manager S. Paul Operations Supervisor
J. Priest Radiation Protection Manager
M. Santiago Supervisor, Operations Training
- LIST [[]]
- OF [[]]
- ITEMS [[]]
- CLOSED [[]]
- AND [[]]
- DISCUS [[]]
SED
Opened and Closed 09-013
TS 3.1 When a Surveillance Requirement Was Not Met
05000293/2008-005-01 NCV Failure to Conduct a Risk Assessment for Emergent
Maintenance on the High Pressure Coolant Injection System
05000293/2008-005-02
RCIC Isolation
Closed
05000293/2008-003-00
- LER [[]]
RCIC System Declared Inoperable During Surveillance Testing Due to Procedure Error
05000293/2007-003-04
TS 4.0.3 When it is Discovered that a Surveillance Has Never Been Performed
A-2
- LIST [[]]
- OF [[]]
- DOCUME [[]]
- NTS [[]]
- REVIEW [[]]
ED Section 1R01 Procedure 8.C.40, Revision 22, Seasonal Weather Surveillance Procedure 2.2.35, Revision 42, Condensate Storage & Transfer System
Procedure 2.1.37, Revision 25, Coastal Storm Preparations and Actions
Procedure 2.1.42, Revision 7, Operation During Severe Weather
Section 1R04 Procedure 2.2.36, Revision 63, Instrument Air Systems
Instrument Air Drawings
Procedure 2.2.23, Revision 32, Automatic Depressurization System
EOP-01, Reactor Pressure Vessel Control
- UFS [[]]
AR Section 10.4, Fuel Pool Cooling and Cleanup System
Procedure 2.2.85, Revision 73, Fuel Pool Cooling and Filtering System
Passport Database Equipment ID Printout on Spent Fuel Pool Pumps Suction Header Drain Valve
Procedure 1.17.1, Revision 9, Potential Seismic Interaction Hazards
Procedure
ID 212, Revision 91, Service Water System
Procedure 2.2.20, Revision 70, Core Spray System
- UFS [[]]
AR Section 6, Core Standby Cooling Systems
- PN [[]]
- P& [[]]
ID M242, Core Spray System
Section 1R05
PNP-2008-03122, Fire Protection Water Pipe Corroded Above Cable Trays
Fire Hazard Analysis, Fire Zone Data Sheet, Fire Area 1.9, Fire Zone 1.9, Reactor Building El. 23' to El. 51'/East Side Fire Protection Engineering Evaluation (FPEE) 49, Revision 0, Barrier Between "A" Division Battery Room and Switchgear Room
- FP [[]]
III-T Penetration Seal in Barrier 194.504A between "B" Switchgear Room and Stairway No. 8
A-3FPEE 123, Revision 0, New Red Line Building - Exposure to Process Buildings
MTS-3 Installation on Enclosure No. 2
Procedure 8.B.17.2, Revision 9, Inspection of Fire Damper Assemblies
Fire Hazards Analysis Fire Zone Data Sheet Fire Area 1.9, Fire Zone 1.13, Fuel Pool Cooling Pumps/Heat Exchanger Area Engineering Evaluation No. 86, Acceptability of Structural Steel Supporting Floor at E1.74 "B" of Reactor Building Procedure 2.2.29, Revision 26, Smoke and Detection Systems
Fire Hazards Analysis Fire Zone Data Sheet, Fire Area 1.9, Fire Zone 3.5, Vital Motor Generator Set Room Exemption Request No. 9, Fixed Fire Suppression with Alternate Shutdown Capability Exemption Request No. 23, Walls with Ratings less than 3 Hours
Engineering Evaluation No. 19, Non-Fire Rated Materials in Seismic Joints
Engineering Evaluation No. 98, Cable Tray Penetration in Appendix "A" Barrier
Procedure 5.5.2, Revision 40, Special Fire Procedure
Section 1R11 Instructional Module, Revision 0, Core Damage/Mitigation Assessment
Severe Accident Guidelines
Severe Accident Guidelines
- LO [[]]
RT Overview - Fall 2008 Presentation Slides
Severe Accident Management Accident Phenomenology Presentation Slides
Section 1R12
- PA [[]]
- PNP -2008-02121, H2O2 Declared Potentially Maintenance Rule (a)(1) Regulatory Guide 1.160, Revision 2, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants
- NUMARC 93-01, Revision 2, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants
PASS System Health Report
Maintenance Rule (a)(1) Action Plan, Post Accident Sampling System
PNP-02469, Standby Gas Treatment Root Valve 31-HO-8 does not fully shut
Procedure
DC-206, Revision 1, Maintenance Rule (a)(1) Process
Procedure
DC-207, Revision 1, Maintenance Rule Periodic Assessment
Functional Failure Determination Form on
- HP [[]]
- HP [[]]
CI System Health Report
Functional Failure Determination Form on
- HP [[]]
CI Undervoltage
K-117 Coil Degradation Apparent Cause Evaluation
K-117 (a)(1) Action Plan (CR-PNP-2008-03113)
PNP-2008-03113, K-117 Air Compressor has exceeded Maintenance Rule Performance Criteria
Section 1R13
A-4Equipment Out Of Service, Risk Assessment Tool Procedure 1.5.22, Revision 11, Risk Assessment Process
Control Room Logs
Procedure 3.M.1-45, Revision 6, Outage Shutdown Risk Assessment
Risk Assessment Review Checklists
- RC [[]]
- HP [[]]
CI failure
Section 1R15
- HP [[]]
CI Suction Line
Isometric Drawing MI00-256-1, High Pressure Coolant Injection
Generic Letter 2008-01, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems
- GE [[]]
- EC [[]]
- HP [[]]
CI Cabling In-Service Aging Needs to be reviewed
Drawing E320, Revision E19, Turbine Building - Area 7, Conduit and Tray Layout
Drawing E318, Revision E13, Turbine Building - Area 6, Conduit and Tray Layout
- PNP -2008-03446, After Change Out of A802 Control Switch, the A802 Breaker would not close from the Control Room
- UFS [[]]
PNP-2007-04801, Restriction Orifices Missing from Recirculation Pump Instrument Lines
Operability Evaluation for
- P& [[]]
ID Recirculation Pump "B" Instrumentation
Calculation
- S& [[]]
SA166, Revision 0, Instrument Line Break Blowdown
Calculation
- S& [[]]
SA167, Revision 0, Reactor Building Response to Instrument Line Break
Calculation No.
RB Secondary Containment (with and without
Restricting Orifice) and with or without
- SGTS. [[]]
PNP-2008-01995, ACB 104B Disconnect Not Seating Properly
Procedure
PNP-2008-03669, Disconnect 102A has a misaligned blade
Section 1R18
PSV-8008
50.59 Screening Form, Change Orientation of PSV-8008
Engineering Change 7768, Temp Mod Required to Document Heat Detection System Disabled on Pre-Action Sprinkler System for the Turbine Bearings Procedure
LI-100, Revision 7, Attachment 9.1, Process Applicability Determination
A-5UFSAR, Chapter 10.8, Revision 26, Fire Protection System Procedure
DC-128, Revision 3, Fire Protection Impact Reviews
Report Number
ER-Q, Updated Fire Hazards Analysis
Procedure 2.2.26, Revision 40, Deluge Sprinklers and Spray Systems
PNP-2008-03619, Control Room Drawing does not Reflect Temporary Modification Installation
Procedure 1.2.4, Revision 45, Operations Performance Assessment Program (OPAP)
- QA Identifies no audits of temporary modifications were performed by Operations during 2008 Section 1R19
- RC [[]]
IC Isolation and Turbine Trip
Procedure
RHR Pump Discharge Check Valve 1001-67A
CK-1001-67A Compliance Package
CK-1001-67A Quality Inspection Plan
Procedure 3.M.4-53, Revision 4, Check Valve Disassembly and Inspection
Vendor Manual V-0442, Velan Gate, Globe, Stop Check Valves
Drawing M132A8, Velan Forged Bolted Bonnet Swing Check Valve
MO-2301-08 Position Lights Out Procedure 3.M.3-51, Revision 26, Electrical Termination Procedure
Procedure
- 8.I. 30, Revision 5, Operability Test for Valve Indicator Light Verification Procedure 8.I.11.11, Revision 9, Reactor Coolant Pressure Boundary Isolation Valve Cold Shutdown Operability Procedure 8.I.1.1, Revision 21, Inservice Pump and Valve Testing Program
WO 5167091201, MO-1001-34A Breaker Testing
Procedure
MOV Diagnostic Test
Procedure
RHR Pump "A" Relay Testing
Procedure
- KV Protective Relay Calibration/Functional Test and Annunciator Verification Drawing E5-200, Revision 7, 4160 Volt Switchgear Relay Settings
PNP-2008-03262, As-Found Data Out of Specification for RHR "A" Relays
Procedure 1.3.34, Revision 15, Attachment 9, Surveillance Test Review for Procedure 3.M.3-1 dated 10/15/2008 Procedure 8.5.2.3, Revision 47, Low Pressure Coolant Injection and Containment Cooling Motor
A-6Operated Valve Operability Test Procedure 8.5.2.2.1, Revision 51, Low Pressure Coolant Injection System Loop "A" Operability - Pump Quarterly and Biennial (Comprehensive) Flow Rate Tests and Valve Tests
SSW P-208D in accordance with 3.M.4-14.2
Procedure 3.M.4-14.2, Revision 53, Salt Service Water Pumps; Routine Maintenance
Drawing M8-39, Revision 1,
- SSW Pump Bearing Retainer for Bronze-Backed Cutless Rubber Standard and Oversized O.D. Drawing M8-38, Revision 7,
SSW Pump Lineshafts
Drawing M8-4, Revision 27, Assembly Drawing Service Water Pump P208A, B, C, D, & E
SSW P-208, Remove/Replace Expansion Joint, Replace 29-CK-3880D with Refurbished Valve, Inspect AV-38006 Procedure 3.M.3-4, Revision 53, Attachment 21, Insulation Test of 480V Related Loads and Cables Procedure 3.M.3-51, Revision 26, Electrical Termination Procedure
Procedure 3.M.3-17.1, Revision 23, Raychem or Taping of 1000 Volt and Under Cables and/or wires Procedure 3.M.1-14, Revision 22, General Maintenance Procedure for Heavy Load Handling Operations Drawing E52A1, Revision E2, Outline and Dimension Salt Service Pump Motor
Drawing M212, Revision 91,
- P& [[]]
ID Service Water System
WO 00154061 03, Post Maintenance Test P-208D
Procedure 8.5.3.2.1, Revision 21, Salt Service Water Pump Quarterly and Biennial (Comprehensive) Operability and Valve Operability Tests Procedure 3.M.1-15, Revision 42, Vibration Monitoring for Preventive Maintenance and Balancing
Procedure
NDE-10.02, Revision 3, VT-2 Examination
WO 00164079, Replace Three Relaying Current Transformers at F15
Schematic Meter and Relay Diagram 23KV Line and Shutdown Transformer
Product Bulletin for Type KOR-15C Current Transformer
Test Results by Omicron for Excitation Curve Data
Section 1R20
Procedure 1.3.37, Revision 27, Post Trip Review
Reactor Plant Event Notification Worksheet
PNP-2008-03963, Following Reactor Scram, Y3 and Y4 buses were de-energized
Y-3/Y-4 Load List/Drawings
PNP-2008-03969, Relief/Safety Valve Leakage Alarms
Control Room Logs
Procedure 5.3.18, Revision 27, Loss of 120V AC Safeguard Buses Y3 and Y31
Procedure 5.3.19, Revision 30, Loss of 120V AC Safeguard Buses Y4 and Y41
A-7Procedure 2.2.12, Revision 38, 120V AC Safeguard Power Supply: Y3-Y4, Y31-Y41, and Y13-Y14 Procedure 2.1.1, Revision 166, Startup from Shutdown
Procedure 2.1.4, Revision 26, Approach to Critical
Forced Outage Schedule
Forced Outage Action Item List
Section 1R22 Procedure 8.4.1, Revision 64, Standby Liquid Control Pump, Quarterly and Biennial Capacity and Flow Rate Test
- SB [[]]
LC Pump pushbutton during testing
Procedure 8.9.1, Revision 111, Emergency Diesel Generator and Associated Emergency
AC Power Source
Section 2OS1
Procedure 6.1-009, Revision 15, Radiological Controls for Handling Highly Radioactive Objectives
and Refuel Floor Activities
Section 2OS2
Procedure
- ALA [[]]
RA Program
Procedure 6.1-220, Revision 4, Radiological Controls for High Risk Evolutions
Pilgrim Station
ALARA Reviews:
07-0066, 07-0078, 07-0079, 07-0080, 07-0081, 07-0116, 07-0140
Section
EDG October 2007 - September 2008 NEI-99-02, Revision 5, Regulatory Assessment Performance Indicator Guidelines
Section
PNP-2007-4936, SRV Insulation
A-8CR-PNP-2007-3432,
PNP-2004-1368, As-Found Setpoint Testing at Wiley Labs
- EE 01-022, Engineering Evaluation for Target Rock Corporation Two-Stage Safety Relief Valve 203- 3C, Rev. 1, dated 04/12/01 Procedure 2.2.23, Revision 32, Automatic Depressurization System
LI-102, Revision 12, Corrective Action Process
Entergy Quality Assurance Program Manual, dated 04/15/08
Automatic Depressurization System Reference Text, Rev. 4
BWROG SRV Leakage Reduction, Rev. 0, Class 1, July 2003
General Electric (GE) Service Information Letter (SIL) 196, Summary of Recommendations for Target Rock Main Steam Safety Relief Valves, dated 09/30/76
- GE [[]]
- SIL 196, Supplement 11, Recommendations Applicable to the Target Rock Main Steam Safety Relief Valve Model #7567F (Two-Stage
- GE [[]]
- 1983 GE [[]]
- GE [[]]
SIL 196, Supplement 16, Target Rock SRV Insulation Maintenance, dated 09/03/92
License Amendment No. 56, Incorporation of
SRV Setpoints, dated 03/26/84
License Amendment No. 222, Deletion of Requirement for
- NRC Regulatory Issue Summary 200-12, Resolution of Generic Safety Issue B-55, "Improved Reliability of Target Rock Safety Relief Valves, dated 08/07/00
- MEMO Closeout of Generic Safety Issue B-55, Improved Reliability of Target Rock Safety Relief Valves (
- NUR [[]]
- NRC [[]]
- NRC [[]]
- NRC [[]]
IN 88-30, Target Rock Two-Stage SRV Setpoint Drift Update, dated 05/25/88
Technical Specifications Section 3.6, Primary System Boundary
- UFS [[]]
- UFS [[]]
AR 6.1 - 6.5, Core Standby Cooling Systems
Vendor Manual No. V0353, Target Rock Safety Relief Valve Model 7567F, dated 07/29/08
Procedure 1.3.34.4, Revision 17, Compensatory Measures
Pilgrim Operator Workarounds Aggregate Report
Compensatory Measures and Disabled Annunciater Logs
Procedure 5.3.35.1, Revision 4, Transient Response Hardcards for Operating Crews
Procedure 2.2.21, Revision 73,
- HP [[]]
CI System
Procedure 2.2.22, Revision 69,
- RC [[]]
PNP-2007-04865, Review of 2007 events
Operations Night Orders
- TE [[]]
AR dated 11/24/2008
A-9Licensing Training for Engineering Staff Procedure
LI-118, Revision 7, Effectiveness Review Criteria
Pilgrim Station Quarterly Trend Report for Third Quarter 2008
Mentorship Pilot Program
PNP-2007-03036, Mispositioning of Exhaust Fan Control Switches
Section 4OA3
Event Notification Sheet No. 44545
Procedure
RCIC Steam Line High Temperature
Control Room Logs
Event Timeline
Plymouth Fire Department Fire Investigation Summary
Incident Report of Emergency Declared at Pilgrim Nuclear Power Station
Radiological Survey Forms
Procedure
IP-100, Revision 29, Emergency Classification and Notification
Root Cause Analysis Report for
PNP-2008-03433, Fire in HP Calibration Room
Fire in HP Calibration Room and Battery Use/Storage Precautions Tailgate Message
Technical Specifications
Emergency Action Level Chart
- RC [[]]
- RC [[]]
IC inoperable due to power supply capacitors age significantly exceeds
recommended limits
- PN [[]]
PS December Forced Outage Startup Chart
Procedure 5.3.18, Revision 27, Loss of 120V AC Safeguard Buses Y3 and Y31
Procedure 5.3.19, Revision 30, Loss of 120V AC Safeguard Buses Y4 and Y41
Procedure 2.2.12, Revision 38, 120V
- AC Safeguard Power Supply: Y3 - Y4, Y31 - Y41, and Y13 - Y14 Procedure 1.3.37, Revision 27, Post-Trip Reviews
- LIST [[]]
- OF [[]]
- ACRONY [[]]
- MS [[]]
- ADA [[]]
- ALA [[]]
RA As Low As Reasonably Achievable
CFR Code of Federal Regulations
CR Condition Report
DRP Division of Reactor Projects
DRS Division of Reactor Safety
EDG Emergency Diesel Generator
A-10EOP Emergency Operating Procedure
- FP [[]]
EE Fire Protection Engineering Evaluation
- HP [[]]
CI High Pressure Coolant Injection
IR Inspection Report
NCV Non-Cited Violation
NEI Nuclear Energy Institute
NRC Nuclear Regulatory Commission
- PA [[]]
- PA [[]]
SS Post Accident Sample System
PI Performance Indicator
- PN [[]]
PS Pilgrim Nuclear Power Station
- RC [[]]
IC Reactor Core Isolation Cooling
RFO Refueling Outage
RWP Radiation Work Permit
- SB [[]]
- SC [[]]
BA Self Contained Breathing Apparatus
SDP Significance Determination Process
SRO Senior Reactor Operator
SSW Salt Service Water
- UFS [[]]