IR 05000293/2007002

Download: ML071270525

Text

May 7, 2007

Mr. Kevin BronsonSite 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 05000293/2007002

Dear Mr. Bronson:

On March 31, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an inspectionat your Pilgrim Nuclear Power Station. The enclosed integrated inspection report documents the inspection findings, which were discussed on April 5, 2007, with you and members of your staff.The inspection examined activities conducted under your license as they relate to safety andcompliance 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.This report documents one self-revealing finding of very low safety significance (Green), whichinvolved a violation of NRC requirements. Additionally, two licensee-identified violations which were determined to be of very low safety significance are listed in this report. However, because of the very low safety significance and because they have been entered into your corrective action program, the NRC is treating these findings as non-cited violations (NCVs),

consistent with Section VI.A.1 of the NRC's Enforcement Policy. If you contest any NCV in this report, you should provide a response with the basis for your denial, within 30 days of the date of this inspection report, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington, D.C. 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C.

20555-0001; and the NRC Resident Inspector at Pilgrim.In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, itsenclosure, and your response (if any) will be available electronically for public inspection in the K. Bronson2NRC Public Document Room or from the Publicly Available Records (PARS) component of theNRC'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/Raymond J. Powell, ChiefProjects Branch 5 Division of Reactor ProjectsDocket No. 50-293License No. DPR-35

Enclosure:

Inspection Report 05000293/2007002

w/Attachment:

Supplemental Informationcc w/encl:G. J. Taylor, Chief Executive Officer, Entergy Operations M. Kansler, President, Entergy Nuclear Operations, Inc.

J. T. Herron, Senior Vice President M. Balduzzi, Senior Vice President, Northeastern Regional Operations C. Schwarz, Vice-President, Operations Support S. J. Bethay, Director, Nuclear Safety Assurance O. Limpias, Vice President, Engineering J. F. McCann, Director, Licensing C. D. Faison, Manager, Licensing R. Patch, Director of Oversight, Entergy Nuclear Operations, Inc.

B. S. Ford, Manager, Licensing, Entergy Nuclear Operations, Inc.

T. C. McCullough, Assistant General Counsel S. Lousteau, Treasury Department, Entergy Services, Inc.

Director, Radiation Control Program, Commonwealth of Massachusetts W. Irwin, Chief, CHP, Radiological Health, Vermont Department of Health The Honorable Therese Murray The Honorable Vincent deMacedo Chairman, Plymouth Board of Selectmen Chairman, Duxbury Board of Selectmen Chairman, Nuclear Matters Committee Plymouth Civil Defense Director D. O'Connor, Massachusetts Secretary of Energy Resources J. Miller, Senior Issues Manager Office of the Commissioner, Massachusetts Department of Environmental Protection Office of the Attorney General, Commonwealth of Massachusetts Electric Power Division, Commonwealth of Massachusetts R. Shadis, New England Coalition Staff D. Katz, Citizens Awareness Network Chairman, Citizens Urging Responsible Energy

SUMMARY OF FINDINGS

...................................................iii

REPORT DETAILS

..........................................................1

REACTOR SAFETY

.........................................................11R01Adverse Weather Protection .......................................1

1R23 Temporary Plant Modifications.....................................12

1EP6Drill Evaluation.................................................12OTHER ACTIVITIES........................................................134OA1Performance Indicator Verification..................................13 4OA2Identification and Resolution of Problems............................144OA3Event Follow-up................................................15 4OA6Meetings, Including Exit..........................................16 4OA7Licensee-Identified Violations.....................................16ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

................................................A-1

LIST OF ITEMS

OPENED, CLOSED AND DISCUSSED...........................A-1

LIST OF DOCUMENTS REVIEWED

..........................................A-1

LIST OF ACRONYMS

......................................................A-6

iiiSUMMARY

OF [[]]

FINDINGSIR 05000293/2007-002; 01/01/2007-03/31/2007; Pilgrim Nuclear Power Station;Post-Maintenance Testing.The report covered a 13-week period of inspection by resident and region-based inspectors. One Green finding, which was a non-cited violation (NCV), was identified. The significance of

most findings is indicated by their color (greater than Green, or Green, White, Yellow, Red)

using

IMC 0609, "Significance Determination Process" (

SDP). Findings for which the SDP does

not apply may be Green or be assigned a severity level after NRC management review. The

NRC 's program for overseeing the safe operation of nuclear power reactors is described in

NUREG -1649, "Reactor Oversight Process," Revision 3, dated July 2000.A.

NRC -Identified and Self-Revealing FindingsGreen. A Green self-revealing

NCV of 10

CFR 50, Appendix B, Criterion XVI,"Corrective Action," was identified for Entergy's failure to promptly correct a condition

adverse to quality associated with the "B" emergency diesel generator (EDG). During

the post overhaul surveillance of the "B"

EDG on January 25, 2007, the "B"

EDG

experienced unexpected load oscillations of approximately 150 kilowatt (kW).

Subsequently, on February 23, 2007, oscillations of greater than 200 kW were seen,

which resulted in the shutdown of the "B" EDG and an entry into a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Technical

Specification (TS) Limiting Condition for Operation (LCO). Entergy corrected the kW

load oscillations by replacing the mechanical portion of the "B" EDG governor. The "B"

EDG was declared operable following successful testing. The issue was entered into

Entergy's corrective action program. The inspector determined that this finding was more than minor because it wasassociated with the Equipment Performance attribute of the Mitigating Systems

cornerstone; and, it affected the cornerstone objective of ensuring the reliability,

availability, and capability of systems that respond to initiating events to prevent

undesirable consequences. A Phase 3 SDP evaluation was necessary due to a

potential for a greater than green finding as indicated in the site specific pre-solved

Phase 2 worksheets. The Phase 3 evaluation concluded that the finding was of very low

safety significance (Green). The inspector also determined that this finding had a

cross-cutting aspect in the area of Problem Identification and Resolution, Corrective

Action Program, in that Entergy personnel failed to thoroughly evaluate the unexpected

kW oscillations. (Section 1R19)B.Licensee-Identified ViolationsTwo violations of very low safety significance, which were identified by the licensee,have been reviewed by the inspector. Corrective actions taken or planned by the

licensee have been entered into the licensee's corrective action program. The violations

and corrective actions are listed in Section

4OA 7 of this report.

REPORT [[]]

DETAILSSummary of Plant StatusPilgrim Nuclear Power Station began the inspection period operating at 100 percent corethermal power. Plant power was reduced to 62 percent on January 26, 2007, to perform power

suppression testing to identify and suppress a suspected fuel defect. The plant was returned to

full power the next day. The plant began end-of-cycle coast down on January 30, 2007. The

plant was shutdown on March 17, 2007, in response to an increase in unidentified drywell

leakage. The plant was restarted on March 19, 2007, after repair of a packing leak on reactor

water cleanup (RWCU) valve MO-1201-85. The plant was operating at 83 percent power in

end-of-cycle coast down at the end of the inspection period.1.REACTOR

SAFET [[]]

YCornerstones: Initiating Events, Barrier Integrity, Mitigating Systems1R01Adverse Weather Protection (71111.01) a.Inspection Scope (2 samples)The inspector reviewed licensee activities to protect plant systems during adverseweather during the periods of January 16-18 and 25-26, 2007 (cold temperatures), and

February 14-15, 2007 (coastal storm). The inspector reviewed Entergy's actions tomitigate the impact of adverse weather on key plant systems. The review focused on

outdoor tanks and on environmental conditions in several buildings, including the intake,

emergency diesel generator (EDG), and station blackout diesel generator (SBODG)

buildings. The plant systems reviewed included the

EDG , the

SBODG, the fire water

system, the demineralized water supply system, and the salt service water (SSW)

system. The references used during this review included station procedures 2.1.37,

"Coastal Storms;" 8.C.40, "Cold Weather Surveillance," Attachment 2; and Section

10.9.3 of the Updated Final Safety Analysis Report (UFSAR). The inspector confirmed

that Entergy was identifying weather related issues and had entered them into the

corrective action program. Additional references used for this review are listed in the

attachment to this report. This inspection activity represents two samples for the onset

of adverse weather. b.FindingsNo findings of significance were identified.1R04Equipment Alignment (71111.04).1Partial System Walkdowns a.Inspection Scope (4 samples) The inspector completed a partial review of the risk significant systems listed below to

determine whether the systems were correctly aligned to perform their designated safety

functions. The reviews occurred during periods when a redundant train or system was

2Enclosureout-of-service for maintenance and/or testing, or following restoration of the system ortrain from maintenance. The position of key valves, breakers, and control switches

required for system operability were verified by field walkdown and/or review of the main

control board indications. To ascertain the required system configuration, the inspector

reviewed plant procedures, system drawings, the

UFS [[]]

AR, and Technical Specifications

(TS). The references used for this review are listed in the attachment to this report.

This inspection activity represents four samples.*"A"

EDG on January 23, 2007;*"A" Residual Heat Removal (

RHR) System on January 5, 2007;

• "A" Core Spray on January 29, 2007; and

• Reactor Core Isolation Cooling (RCIC) System on February 21, 2007. b.FindingsNo findings of significance were identified..2Complete System Walkdowns a.Inspection Scope (1 sample)The inspector performed a full system review of the "A"

SSW Loop, including the "C"

SSW Pump, on February 11, 2007, to verify the system was properly aligned and

capable of performing its safety function. To ascertain the required system

configuration, the inspector reviewed plant procedures, system drawings, the

UFS [[]]

AR,

and the TS. The references used for this review are listed in the attachment to this

report. A walkdown of the accessible portions of the system was performed to assess

the system's material condition and the following attributes: *valves were correctly positioned and did not exhibit leakage that would impactthe function(s) of any given valve;*electrical power was available and properly aligned;

• major system components were properly labeled;

• hangers and supports were correctly installed and functional;

• ancillary equipment or debris did not interfere with system performance; and

• valves were locked as required by the locked valve program.This inspection activity represents one sample. b.FindingsNo findings of significance were identified.

3Enclosure1R05Fire Protection (71111.05).1Quarterly Fire Protection Inspection a.Inspection Scope (12 samples)The inspector toured selected areas of the plant to observe conditions related to: (1) transient combustibles and ignition sources; (2) fire detection systems; (3) manual

firefighting equipment and capability; and (4) passive fire protection features. The

inspector reviewed the material condition of active and passive fire protection system

features, and their operational lineup and readiness. The inspector also reviewed the

applicable fire hazard analysis fire zone data sheets. The inspector verified that the

licensee addressed fire protection deficiencies in the corrective action program. The

references used for this review are listed in the attachment to this report. This

inspection activity represents twelve samples.*UPS [Uninterruptible Power Supply] Diesel Building;*Fire Zone 1.3, Reactor Building 17 ft, High Pressure Coolant Injection (HPCI)Pump/Turbine Room;*Fire Zone 5.6, Electric Fire Pump and Open Areas of Intake, and Yard Areas;

• Fire Zone 3.2, Radwaste & Control Building 23 ft, Cable Spreading Room;

• Fire Zone 1.5, Reactor Building,

RC [[]]

IC Room;

• Fire Zone 1.1, Reactor Building, "A" RHR and Core Spray Room;

• Fire Zone 1.6, Reactor Building, Control Rod Drive Pump Room;

• Fire Zone 2.1, Radwaste & Control Building 23 ft, "B" Switchgear and LoadCenter Room;*Fire Zone 1.9A, "A" RHR Valve Room;

• Fire Zone 1.2, "B" RHR & Core Spray Quadrant;

• Fire Zone 1.4,

HP [[]]

CI Control Panel; and

• Fire Zone 1.7,

RC [[]]

IC Mezzanine. b. FindingsNo findings of significance were identified..2Annual Fire Drill Observation a.Inspection Scope (2 samples)The inspector observed two unannounced fire drills:

• Fire in a cable tray, Reactor Building 23 ft elevation; and*Fire in the elevator machine room, Operations and Maintenance (O&M) Building.The fire drills were conducted in accordance with plant procedure

ENN -

DC-189, "FireDrills." The inspector observed performance of the fire brigade personnel, and

confirmed that the licensee's fire fighting pre-plan strategies were utilized, the

4Enclosurepre-planned drill scenario was followed, and the drill objectives were met. The inspectorconfirmed that, as appropriate, proper security and radiological controls were applied;

proper protective clothing and breathing apparatus were donned; sufficient fire fighting

equipment was brought to the scene; the fire brigade leader's fire fighting directions

were clear; and communications with the plant operators and between fire brigade

members were effective. The inspector confirmed the drill critique identified areas to

enhance fire brigade performance. The inspector verified that the licensee identified

appropriate corrective actions for identified deficiencies and entered the issues into the

corrective action program. This inspection activity represents two samples. b.FindingsNo findings of significance were identified.1R06Flood Protection Measures (71111.06) a.Inspection Scope (1 sample)On January 31, 2007, the inspector reviewed protective measures in place to protectagainst internal flooding of the "A" and "B" EDG rooms. The inspector performed visual

inspections of the water scuppers on the perimeter of the EDG building to determine

whether there were obstructions. Building floor drains were inspected to determine

whether there were blockages. Flood barriers which separate the "A" and "B" EDG

compressor rooms were inspected to ensure they could perform their intended function.

This inspection activity represents one sample for internal flood protection. b.FindingsNo findings of significance were identified1R11Licensed Operator Requalification Program (71111.11)Resident Inspector Quarterly Review a.Inspection Scope (1 sample)The inspector observed a licensed operator simulator exam given on January 23, 2007. The exam was administered using scenario SES-047, Revision 1, and involved both

operational transients and design basis events. The inspector verified that simulator

conditions were consistent with the scenario and reflected the actual plant configuration

(i.e., simulator fidelity). The inspector observed the crew's performance to determine

whether the crew met the scenario objectives, accomplished the critical tasks,

demonstrated proper use of abnormal and emergency operating procedures,

demonstrated proper command and control, and communicated effectively. The

inspector observed the evaluators' post-scenario critique and confirmed items for

improvement were identified and discussed with the operators to further enhance

performance. This inspection activity represents one sample.

5Enclosure b.FindingsNo findings of significance were identified.1R12Maintenance Effectiveness (71111.12) a.Inspection Scope (3 samples)The inspector reviewed the follow-up actions for selected system, structure, orcomponent (SSC) issues and reviewed the performance history of these SSCs to

assess the effectiveness of Entergy's maintenance activities. The inspector reviewed

Entergy's corrective actions for these issues in accordance with Pilgrim procedures and

the requirements of 10 CFR 50.65(a)(1) and (a)(2), "Requirements for Monitoring the

Effectiveness of Maintenance." In addition, the inspector reviewed selected SSC

classification, performance criteria and goals, system health reports, and corrective

actions that were taken or planned to verify whether the actions were reasonable and

appropriate. The inspector attended licensee meetings and reviewed licensee plans to

address the systems in maintenance rule (a)(1) status. The following issues were

reviewed:*Classification of equipment failures for System 48, the standby gas treatment(SBGT) system. The inspector reviewed licensee actions for condition reports

(CRs) pertaining to the

SBGT System Instrument Air System, including

CRs

200402377, 200500784, 200501208, 200501250, and 200602645. The

inspector reviewed the licensee's basis for placing the system in maintenance

rule (a)(2) status. *Classification of equipment failures for System 61, the

SBODG. The inspectorreviewed licensee actions for select condition reports pertaining to the

SBODG

system, including CR 20070084. The inspector reviewed the licensee's plans for

returning the system to (a)(2) status by October 2007.*Classification of equipment failures for System 61, the station

EDG. Theinspector reviewed licensee actions for select condition reports pertaining to the

EDG system, including

CRs 20070056, 20070282, 20070552 and 20070703.

The inspector discussed with licensee staff the plans for returning the system to

(a)(2) status by September 2007.This inspection activity represents three samples. b.FindingsNo findings of significance were identified.

6Enclosure1R13Maintenance Risk Assessments and Emergent Work Control (71111.13) a.Inspection Scope (7 samples)The inspector evaluated on-line risk management for planned and emergent work. Theinspector reviewed maintenance risk evaluations, work schedules, recent corrective

actions, and control room logs to verify that concurrent planned and emergent

maintenance or surveillance activities did not adversely affect the plant risk already

incurred with the out-of-service components. The inspector evaluated whether Entergy

took the necessary steps to control work activities, minimize the probability of initiating

events, and maintain the functional capability of mitigating systems. The inspector

assessed Entergy's risk management actions during plant walkdowns. The inspector

also discussed the risk management with maintenance, engineering, and operations

personnel, as applicable, for the activities. References used for the inspection are

identified in the attachment to this report. The inspection activity represents seven

samples. *Planned valve diagnostics on Reactor Building Closed Cooling Water (RBCCW)valve 4060B on January 16, 2007;*Emergent work on a horizontal support insulator for the offsite 345 kV [kilovolt]Line 355 on January 18, 2007;*Emergent work on Reactor Building isolation valve

AO -N-91 onFebruary 8, 2007; *Planned calibration of 345 kV watthour transducers - control room telemeteringinstruments on January 26, 2007;*"B"

EDG emergent work on January 4, 2007;

• "B" EDG governor replacement emergent work February 23, 2007; and

• "A"

EDG [[emergent work on March 14, 2007. b.FindingsNo findings of significance were identified.1R15Operability Evaluations (71111.15) a.Inspection Scope (4 samples)The inspector reviewed selected operability determinations to assess the adequacy ofthe evaluations, the use and control of compensatory measures, compliance with the]]

TS , and the risk significance of the issues. The inspector used the

TS ,

UFS [[]]

AR,

associated design basis documents, and the additional references listed in the

attachment to this report. This inspection activity represents four samples.*CR 20070108, Reasonable Expectation for Operability (REO) for DrywellLeakage Measurement System;*CRs 20070293, 20070703, "B" EDG Load Swings - Operability Evaluation;

7Enclosure*CR 20070629, B18 Motor Control Center and

VRC -203A Air Conditioning Unit;and*

CR 20070670, Dual Indication on

HPCI Steam Isolation Valve

MO -2301-3. b.FindingsNo findings of significance were identified.

NRC review of the "B"

EDG is discussedfurther in Section 1R19 below.1R17Permanent Plant Modifications (71111.17) a.Inspection Scope (1 sample)The inspector selected a risk-significant plant modification package for review to verifythat the design bases, licensing bases, and performance capability of the risk significantsystem had not been degraded through the modification. The modification selected for

review was Engineering Request 03118068, which restored automatic pumping of the

drywell sumps.For the selected modification, the inspector reviewed the design assumptions andvalidations to determine the design adequacy. In addition, the inspector reviewed the

associated 10 CFR 50.59 safety evaluation to verify that the safety issue pertinent to the

changes was properly resolved or adequately addressed. The inspector also reviewed:

(1) field implementation of the changes to the drywell sump pump controls;

(2) post-modification functional testing to determine the readiness for operations; and,

(3) compensatory measures used to monitor reactor coolant system leakage. The

inspector reviewed the associated drawings to independently verify the changes and

post-work test methods were appropriate. The inspector walked-down portions of the

modification on radwaste panel C20 and observed the status of indications on control

room panels C903 and C904. The inspector monitored the performance of the drywell

leakage system during periodic reviews of plant operations. The inspector reviewed the

changes to procedure 2.2.125.1, "Reset of Primary and Secondary Containment

Isolations (Group I,

II ,

III ,

IV , V,

VI and VII)," which were to enhance the administrativecontrol of the sump pumps and isolation valves after a containment isolation.

References used during this review as listed in the attachment to this report. This

inspection activity represents one sample. b.FindingsNo findings of significance were identified.1R19Post-Maintenance Testing (71111.19) a.Inspection Scope (5 samples)The inspector reviewed post-maintenance test (PMT) activities on risk significantsystems to determine whether the effect of the test on the plant had been evaluated

8Enclosureadequately, the test was performed in accordance with procedures, the test data metthe required acceptance criteria, and the test activity was adequate to verify system

operability and functional capability following maintenance. The inspector confirmed

that systems were properly restored following testing and that discrepancies were

appropriately documented in the corrective action process. References used during this

review are listed in the attachment to this report. The inspection activity represents five

samples.*MR 02119435 and

MR 02119532, Maintenance and Inspection of

RHR valvesMO-1301-26 and

MO -1301-62;*

MR 07102225, Post Work Test for Refuel Floor Isolation Valve AO-N-91;

• MR 04116408, 2-Year Preventive Maintenance Overhaul of "B" EDG;

• MR 07101429, Replacement of

EGB Governor Special Test, "B"

EDG; and

• MR 07104403,

RWCU [[]]

MO -1201-85 packing replacement and Operability

PMT per 8.6.5.2. b.FindingsIntroduction. A Green self-revealing

NCV of

10 CFR 50, Appendix B, Criterion

XVI,"Corrective Action," was identified when Entergy failed to promptly correct a condition

adverse to quality associated with the "B"

EDG. As a result, the

EDG was not capable

of performing its safety function for 29 days. Description. On January 21, 2007, the "B" EDG was taken out of service for a plannedmaintenance overhaul. The overhaul consisted of inspections and preventive

maintenance of the "B" EDG and support systems. One of the maintenance tasks was

a flush and change of the "B" EDG governor oil. On January 25, 2007, following the

completion of the overhaul, Entergy began a "B" EDG operability test which included a

full load run to 2600 kW, in accordance with Entergy procedure 8.9.1, "Emergency

Diesel Generator and Associated Emergency Bus Surveillance." During the operability

run, operators noted unexpected oscillations in both kilowatts and generator amperage.

Kilowatt oscillations measured approximately 150 kW and generator amperage

oscillations measured approximately 10 to 12 amperes. Procedure 8.9.1 states that if

kilowatt oscillations are greater than 200 kW, then shutdown of the diesel engine is

required. Although not exceeding the 200 kW abort criteria, Entergy entered the

degraded condition in their corrective action program as CR 20070293. Entergy

concluded that the oscillations were minor and, since they did not exceed the abort

criteria in procedure 8.9.1, the operability run was successful and the "B" EDG was

operable. Corrective actions planned for the kW oscillations included instrumenting the

"B" EDG with a test recorder to obtain data to tune the governor in order to achieve a

more stable operation of the "B" EDG. The corrective actions were scheduled to

coincide with the next surveillance run on February 23, 2007, which was 29 days after

the initial kW oscillations were observed.During the February 23, 2007, "B" EDG operability run, Entergy again noted kWoscillations at a full load of 2600 kW. Entergy noted this condition in their corrective

action program in CR 20070703. Specifically, approximately 50 minutes into the full

9Enclosureload portion of the operability run, kW oscillations were observed at 100 kW andcontinued to increase during the run to 200 kW at a full load of 2600 KW. The full load

run portion of the test was completed and the control room operator started to unload

the engine in accordance with the procedure; as the load was reduced to 1800 kW, the

kW oscillations were observed to exceed 200 kW. At this time, the control room

supervisor ordered that the "B" EDG be secured due to the procedure abort criteria of

greater than 200 kW being exceeded. Entergy's subsequent investigation included

reviewing the test recorder data and sampling the oil from the mechanical portion of the

governor (referred to as the "EGB"). Entergy found that the sampled oil from the EGB

was gray and cloudy, and had a burnt smell, unlike the oil that was installed in the

governor during the January 2007 overhaul. Initial testing of the EGB oil by the site

Chemistry department indicated the presence of unidentified contaminants. Based

largely on the condition of the governor oil sample, Entergy concluded that the EGB

required replacement. Entergy Maintenance replaced the EGB on February 24, 2007, and tuned the governor

system in accordance with Entergy procedure 3.M.3-61.7, "Woodward Governor

Tuning." The governor oil system was vented, and on February 25, 2007, an operability

run was commenced. During the operability run, load oscillations did not exceed 30 kW.

Entergy further tested the governor by performing a load-reject test of the "B" EDG. The

purpose of this test was to monitor the response of the governor control system when a

large load (core spray pump) was started and tripped. This test was satisfactorily

completed. The "B"

EDG was then declared operable.On March 7, 2007, Entergy shipped the

EGB to the vendor for testing and failureanalysis. When tested by the vendor, the EGB lost pressure within a few minutes and

the test was aborted. The vendor performed an inspection of the EGB and found shreds

of aluminum in the governor oil. The source of the aluminum was determined to be an

aluminum label that was originally attached to the shutdown solenoid valve located

inside the EGB. Entergy concluded that the most likely scenario was that the label was

installed during maintenance in 2002 and became dislodged as a result of the governor

oil change during the January 2007 engine overhaul. The label was then ground up by

the relay bushing gears. The ground particles then clogged the internal passages of the

governor, resulting in a degraded governing ability of the EGB. This resulted in the

unexpected kW oscillations of the "B"

EDG.E ntergy conducted an extent of condition review and tested the oil from the "A"

EDGgovernor. Test results showed that the oil in the "A" EDG governor was acceptable with

no aluminum particles present. The root cause investigation concluded that this issue

was isolated to the "B" EDG. Entergy analyzed the effects that the 200 kW oscillations

would have had on connected emergency core cooling system loads while the "B" EDG

operated. The frequency changes resulting from 200 kW oscillations were found not to

cause any premature tripping of loads or other adverse consequences.Analysis. The inspector determined that the failure to promptly correct a conditionadverse to quality associated with the "B" EDG was a performance deficiency. While

the kW swings during the January test did not exceed the 8.9.1 abort criteria, the

10Enclosureoscillations were not typical of historical performance. The inspector concluded that the"B"

EDG was inoperable for 29 days because the

EGB function was impacted with an

ongoing degradation mechanism. The inspector determined that this finding was more

than minor because it was associated with the Equipment Performance attribute of the

Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring

the reliability, availability, and capability of systems that respond to initiating events to

prevent undesirable consequences.The finding was determined to be of very low safety significance (Green) in accordancewith Inspection Manual Chapter (IMC) 0609, Appendix A, "Determining the Significance

of Reactor Inspection Findings for At-Power Situations." The inspector initially evaluated

the significance of this finding using IMC 0609, Appendix A, and determined that a

Phase 2 analysis was required since the finding represented an actual loss of safety

function of a single train for greater than the

TS allowed outage time. An

NRC Senior

Reactor Analyst (SRA) determined a Phase 3 analysis was necessary due to the

indication of a greater-than-green finding using the site-specific Phase 2 notebook. The

SRA used the Pilgrim Standardized Plant Analysis Risk (

SPAR) model to analyze the

finding and noted differences in how the Phase 2 notebook and the

SP [[]]

AR model

credited the use of the 23kV offsite power source and the

SBODG. The

SPAR model

dominant cutsets are a station blackout with the

HP [[]]

CI train unavailable and a stuck

open safety relief valve. The

SRA determined that the

SPAR model correctly

characterized the significance of the finding. The SRA further evaluated the potential

risk contribution from large early release frequency and the contribution from external

events including seismic and fire initiated loss of offsite power events. Based on this

review, the SRA concluded that the finding was of very low safety significance (Green).The inspector determined that this finding had a cross-cutting aspect in the area ofProblem Identification and Resolution, Corrective Action Program, in that Entergy

personnel failed to thoroughly evaluate the unexpected kW oscillations.Enforcement.

10 CFR 50, Appendix B, Criterion

XVI, "Corrective Action," requires thatmeasures be established to assure that conditions adverse to quality are promptly

identified and corrected. Contrary to the above, on January 25, 2007, Entergy failed to

promptly correct a condition adverse to quality when load swings of 150 kW were found

during a post maintenance operability run of the "B" EDG. Because this violation is of

very low safety significance and has been entered in Entergy's corrective action

program (CR 20070703), this violation is being treated as an

NCV , consistent withSection

VI.A. 1 of the

NRC Enforcement Policy:

NCV 05000293/20070201, InadequateEvaluation of Unexpected Emergency Diesel Generator Load Swings.

11Enclosure1R20Refueling and Other Outage Activities (71111.20) a.Inspection Scope (2 samples)The inspector reviewed Entergy's plans and preparations for the refueling outagescheduled to begin in April 2007. This inspection activity represents two samples.*Review of Refuel Outage Plan. The inspector reviewed the refueling outage(RFO) -16 Outage Shutdown Risk Assessment and procedure TP07-023,

"Compensatory Measures," to verify that Entergy addressed the outage impact

on defense-in-depth for the five shutdown critical safety functions: electrical

power availability, inventory control, decay heat removal, reactivity control, and

containment. The inspector reviewed how Entergy planned to provide adequate

defense-in-depth for each safety function, and the planned contingencies to

minimize the overall risk where redundancy was limited or not available.

Consideration of operational experience was also assessed. *New Fuel Receipt and Inspections. The inspector observed licensee activities toreceive and inspect new fuel for Operating Cycle 17, install fuel channels, and

store the fuel in the spent fuel pool. The inspector used the following references

for the review: Procedure 4.1, "Receiving and Handling of Unirradiated Fuel

Assemblies;" Procedure 4.2, "Inspection and Channeling of Nuclear Fuel;"

TS 3.7, "Containment Systems;" and

UFSAR [[Section 10.3, "Spent Fuel Storage." b.FindingsNo findings of significance were identified. 1R22Surveillance Testing (71111.22) a.Inspection Scope (8 samples)The inspector observed surveillance tests and/or reviewed test results to determinewhether the test acceptance criteria were consistent with]]

TS, that the tests were

performed in accordance with the written procedure, that the test data was complete

and met procedural requirements, and the components were capable of performing their

intended safety functions. Additional references used for this review are listed in the

attachment to this report. The inspection activity represents eight samples.*2.1.15,

RCS [[[Reactor Coolant System] Leak Rate determination per]]

TS 3.6.C;*8.5.5.9,

RCIC Simulated Automatic Actuation, Flow Rate, and Cold QuickstartTest;*8.9.1, "A"

EDG Monthly Surveillance;

2Enclosure*8.M.2-1.5.8.3, Logic System Functional Test of System A Standby GasTreatment Initiation, Reactor Building Isolation and Inboard Drywell Isolation

Valves (Atmospheric Control Valves);*7.3.36, Offgas Sampling and Air Ejector Radiation Monitor Setpoints;*8.5.3.1,

RBCCW System Quarterly and Biennial Comprehensive Operability(P-202D);*8.5.1.1, Core Spray Pump Biennial

IST [Inservice Test]; and*8.A.18, Core Spray System Integrity Surveillance. b.FindingsNo Findings of significance were identified. Licensee identified findings are described inSection 4OA7 of this report.1R23Temporary Plant Modifications (71111.23) a.Inspection Scope (2 samples)The inspector reviewed the temporary modifications identified below to verify that thelicensing bases and performance capability of the associated risk significant system had

not been degraded through the modification. A walkdown was performed to determine

whether equipment was installed in accordance with instructions. The inspector

reviewed applicable drawings and procedures to determine whether they properly

reflected the temporary modifications. The references used for this review are listed in

the attachment to this report. This inspection activity represents two samples.*Temporary Alteration 06-1-064, Drill Through Disc of 29-HO-3A; and*Temporary Alteration 06-1-063, Plexiglass in Screen House. b.FindingsNo findings of significance were identified.Cornerstone: Emergency Preparedness1EP6Drill Evaluation (71114.06) a.Inspection Scope (1 sample) The inspector observed an evaluated licensed operator simulator training exercise onJanuary 23, 2007, and evaluated the crew's ability to implement the emergency plan.

Specifically, the inspector confirmed that the crew properly classified the event,

13Enclosureactivated the notification system, and appropriately completed and transmitted the eventnotification forms in a timely manner. This inspection activity represents one sample. b.FindingsNo findings of significance were identified.4.OTHER

ACTIVI [[]]

TIES [OA]4OA1Performance Indicator Verification (71151).1Reactor Safety Cornerstones a.Inspection Scope (3 samples)The inspector reviewed Performance Indicator data to determine the accuracy andcompleteness of the reported data. The review was accomplished by comparing

reported Performance Indicator data to confirmatory plant records and data available in

plant logs, the chemistry data base, maintenance rule records, Licensee Event Reports,

condition reports, and NRC inspection reports. The inspection activity represents three

samples.*Barrier Integrity Cornerstone, Reactor Coolant System Unidentified Leakagefrom the second quarter of 2006 through the first quarter 2007;*Barrier Integrity Cornerstone, Reactor Coolant System Specific Activity from thesecond quarter of 2006 through the first quarter 2007; and*Mitigating System Cornerstone, Safety System Functional Failures from the firstquarter of 2006 through the first quarter 2007. b. FindingsNo findings of significance were identified.

14Enclosure4OA2Identification and Resolution of Problems (71152)Reactor Safety Cornerstone.1Review of Items Entered into the Corrective Action Program a.Inspection ScopeAs required by Inspection Procedure 71152, "Identification and Resolution of Problems," the inspector performed a screening of each item entered into the licensee's corrective

action program. This review was accomplished by reviewing printouts of each condition

report, 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.FindingsNo findings of significance were identified. .2Annual Sample - Station Blackout Diesel Generator a.Inspection Scope (1 sample)The inspector reviewed Entergy's corrective actions for the failure of the

SBO [[]]

DG to starton demand in January 2005. Specifically, on January 23, 2005, following the loss of

electrical power to its auxiliary equipment, operations attempted to start the

SBO [[]]

DG, as

specified in procedure 2.4.146, "Station Blackout Diesel Generator," when jacket water

outlet temperature dropped to

80F. The diesel generator failed to start, tripping on lowlube oil pressure. The inspector reviewed responses to

CRs 200501177, 200500392,

and 200500256; and the procedure changes specified for procedures 2.2.146, "Station

Blackout Diesel Generator;" 2.4.16, "Distribution Alignment Electrical System

Malfunctions;" and 2.4.A.23, "Loss/Degradation of 23 kV Line." b.Findings and ObservationsNo findings of significance were identified. However, a negative observation regardingEngineering Department performance was noted during this review. Engineering's

original proposed solution, to increase the jacket water outlet temperature from 80F to85F, lacked an adequate basis to assure the diesel engine would start. Afterprompting by Operations Department personnel, the issue was re-evaluated and

Entergy determined that the appropriate corrective action was the revision of station

procedures to require starting the

SBO [[]]

DG as soon as possible following a loss of power

to its auxiliary equipment. Prior testing had demonstrated the

SBO [[]]

DG was able to start

15Enclosure10 minutes following loss of power to its auxiliary equipment. This inspection activityrepresents one sample.4OA3Event Follow-up (71153) a.Inspection Scope (4 samples)The inspector assessed the control room operator performance during the followingplanned and un-planned non-routine evolutions. The inspector evaluated personnel

performance based on control room observations, interviews, and reviews of operator

logs, alarm response procedures, and operating procedures. This inspection activity

represents four samples.*The inspector observed a planned plant power reduction to 62 percent power onJanuary 26-27, 2007, per procedure 2.1.14, "Station Power Changes," and

power suppression testing, per procedure 9.32, "Power Suppression Testing."

The inspector used power maneuvering plan MAN.C16-76 as a reference for this

review. The inspector reviewed Entergy's actions to identify a potential fuel leak,

suppress the fuel defect, and follow the fuel pre-conditioning operating

guidelines.*The inspector reviewed the operator response to a partial loss of the 23 kVpower supply on January 26, 2007, due to a line failure on the site access road.

The inspector reviewed the operator actions per procedure 2.4.A.23,

"Loss/Degradation of 23 kV Line," and the actions to restore plant conditions

following the restoration of power. *The inspector responded to the site on March 17, 2007, to review Entergy'sactions to shutdown and cooldown the plant in response to increasing drywell

leakage. During plant operations at 86 percent power in end-of-cycle coast

down on March 17, 2007, the operators noted unidentified drywell leakage

increased from 0.57 gallons per minute (gpm) to 1.16 gpm over a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />

period. The operators began a controlled shutdown at 12:30 p.m. to place the

plant in cold shutdown in order to investigate and repair the leakage source. The

operators reduced reactor power using recirculation flow and control rods, and

manually scrammed the reactor from 30 percent power at 4:55 p.m. on

March 17, 2007. The reactor responded as expected during the scram and

safety systems operated properly. Reactor Building and containment isolations

and

RW [[]]

CU system isolation occurred as expected due to low reactor vessel

level following the scram [Event Number 43235]. Drywell unidentified leak rate

increased to a maximum value of 2.59 gpm, and then decreased as the reactor

was cooled down and depressurized. Drywell leakage remained below the TS 3.6.C.1 limits for total (25 gpm) and unidentified (5 gpm) leak rate.

16EnclosureThe plant entered cold shutdown at 4:32 pm on March 18, 2007. Entergypersonnel entered the drywell and identified a packing leak from

RW [[]]

CU system

valve

MO -1201-85, which was subsequently repaired (

MR 07104403). This item

was entered into Entergy's corrective action program as CR 20070949 to

investigate the root cause of the packing leak. During the above sequence of events, the inspector observed the operatorsreduce power per power maneuvering plan MAN.C16-84 and shutdown the

reactor by inserting a manual scram per procedure 2.1.6, "Reactor Scram." The

inspector observed Entergy actions to implement plant cooldown activities using

procedure 2.1.5, "Controlled Shutdown from Power," and 2.1.7, "Vessel Heatup

and Cooldown." The inspector reviewed Entergy's scram report and verified that

the cause of the shutdown was understood and corrected prior to plant restart.

Additional references used during this review are listed in the attachment to this

report.*The inspector reviewed Entergy actions on March 18-19, 2007, to restart theplant using procedures 2.1.1, "Startup from Shutdown," and 2.2.70, "Drywell and

Torus Inerting," and power maneuvering plan MAN.C16-84. The startup

activities observed included the approach to criticality, rolling the main turbine,

and synchronizing the unit to the electrical grid. The reactor was restarted at

3:08 p.m. on March 19, 2007. The unidentified leak rate in the drywell was 0.03

gpm following the resumption of plant power operations. The inspector reviewed

Entergy actions to enter deficiencies into the corrective action program. The

references used during this review are listed in the attachment to this report. b.FindingsNo findings of significance were identified.4OA6Meetings, Including ExitExit Meeting SummaryOn April 5, 2007, the inspection results were presented to Mr. Kevin Bronson, andmembers of his staff. The inspector asked the licensee whether any of the material

examined during the inspection should be considered proprietary. No proprietary

information was identified.4OA7Licensee-Identified ViolationsThe following violations of very low safety significance (Green) were identified byEntergy and are violations of

NRC requirements which meet the criteria of Section

VI of

the

NRC Enforcement Policy,

NUREG-1600, for being dispositioned as a Non-Cited

Violation.

17Enclosure*TS 5.4.1 requires, in part, that Entergy establish and implement procedures tocover the activities specified in Regulatory Guide 1.33, Appendix A. Contrary to

the above, the requirements contained in procedures 8.5.3.1, "Reactor Building

Closed Cooling Water System Quarterly and Biennial Comprehensive

Operability," and 1.3.34.2, "Limiting Conditions For Operation Log," were not

implemented on February 24, 2007, for a failed surveillance test of the "D"

RBC [[]]

CW pump (P-202D). Specifically, operators did not identify the pump's total

dynamic head of 81.5 ft was below the minimum allowed total dynamic head of

ft. The error was identified by system engineering on February 26, during

review of the completed procedure. Condition Reports 200700719 and

200700737 were issued and operations entered P-202D into tracking LCO

(Track-1-07-0016), as required by procedure 1.3.34.2. The finding is more than

minor per MC-0612, Appendix E, examples 2a and 2c, in that the total dynamic

head was below the minimum specified value. The issue is not greater than

Green (very low safety significance) because the

RBC [[]]

CW loop remained

operable due to the availability of the two remaining pumps.*TS 5.4.1 requires, in part, that procedures be established and implementedcovering the activities specified in Regulatory Guide 1.33, Appendix A. Contrary

to the above, plant personnel did not implement the requirements of procedures

7.3.36, "Offgas Sampling and Analysis," and 7.4.63, "Process Radiation

Monitoring Setpoints," when samples of the offgas system indicated the

setpoints for offgas radiation monitors 1705-3A/B were non-conservative. The

setpoints are used to assure the controls of Offsite Dose Calculation Manual

(ODCM) 3.1.2 are met. The discrepancy was identified after 26 samples had

been taken from January 16, 2007, through March 3, 2007. The licensee

entered action statement

LCO -

ACT-1-07-035 and completed actions to adjust

the setpoints on March 3, 2007. The issue was more than minor per MC-0612,

Appendix A, because the non-conservative setpoints impact the public radiation

safety cornerstone which involves radiological effluent monitoring and was

contrary to the controls in the

OD [[]]

CM 3.1.2. The issue was not more than Green

because offgas release rates remained well below the levels specified in TS 3.8.1, and other radiation monitors (post-treat and main stack) in the effluent

pathway were operable and would have alerted the operators of the need for

action. This discrepancy was entered into the Entergy's corrective action system

as

CR 20070784.

ATTACH MENT:

SUPPLE [[]]

MENTAL [[]]

INFORM [[]]

ATION A-1AttachmentATTACHMENTSUPPLEMENTAL

INFORM [[]]

ATIONK EY

POINTS [[]]

OF [[]]

CONTAC [[]]

TLicensee personnel:K. BronsonSite Vice President, PilgrimB. CobbMaintenance Supervisor

T. CollisSystem Engineer

W. CookSupervisor Electrical Engineering

F. DicristofaraSenior Operations Specialist

D. EllisSenior Engineer

J. FitzsimmonsRadiation Protection Supervisor

B. FordLicensing Manager

G. JamesReactor Engineering Superintendent

J. KeenanSystem Engineer

T. McElhinneyChemistry Superintendent

C. McMorrowSenior Operations Instructor

J. MoylanElectrical Supervisor

D. NoyesAssistant Operations Manager

E. OlsonOperations Manager

M. SantiagoTraining Superintendent

K. SejkoraSenior Chemistry Specialist

R. SmithGeneral Manager-Plant Operations

NRC personnel:W. Raymond, Senior Resident InspectorC. Welch, Resident InspectorLIST

OF [[]]

ITEMS [[]]

OPENED ,

CLOSED [[]]

AND [[]]

DISCUS SEDOpen and Closed05000293/20070201NCVFailure to Thoroughly Evaluate Degraded Condition on "B"EDG Following January Overhaul (Section 1R19)LIST

OF [[]]

DOCUME NTS

REVIEW [[]]

EDSection 1R01CRs 20070309, 20070310, 20070307, 20070189, 20070162, 20070312, 20070342, 20070315,20070567, 20070188ER 07101779

A-2AttachmentWork Request Tag 088161 Unit Heater ControlMaintenance Requests 07101562, 06114952Section 1R042.2.20, Core Spray, Revision 672.2.19,

RHR , Revision 93P&

ID M241, Residual Heat Removal System, Revision

83P&ID M212, Service Water System, Revision 88Section 1R05

CR s 20070267, 20070676,

20070089ENN -

DC -161, Transient Combustible Program, Revision

189XM -1-

ER -Q, Updated Fire Hazards Analysis, Revision E7ENN-DC-189, Fire Drill Scenario, Revision 05.5.2, Special Fire Fighting Procedure, Revision 36 Special Fire Procedure, Attachment 36, "Security

UPS Diesel Total Flooding

CO2 System,"Revision 36Section 1R06UFSAR 2.4.4, Storm Flooding Protection

Section 1R12CRs 200402377, 200500784, 200501208, 200501250, 200602645, 20070084, 200602122,20070056, 20070282, 20070293, 20070552, 200602626, 20070703(a)(1) Action Plan for Standby Liquid Control SystemAction Plan Number

SENG -

APL -07-001System Health Report - System 61 Station Blackout Diesel GeneratorSystem Health Report - System 61 Station Emergency Diesel GeneratorMaintenance Rule (a)(1) Systems Status and Evaluations PlansMaintenance Rule

SSC Basis Document -

EDG s,

SBO [[]]

DG, Fuel Storage and TransferMaintenance History for System 61 Emergency Diesel Generators for 2005 - 2007Maintenance Requests 06114362, 061165089, 06115255, 06116087, 06116300, 06116452,06116451, 06116450, 06116664, 06116977, 06116981, 06117195, 06118479,

07100895, 07101429, 07102212System

61 LCO -

ACT -1-06-0032,

ACT -1-06-0050,

ACT -1-06-0066,

ACT -1-6-0131,

ACT -1-06-0159, Act-1-07-001,

ACT -1-07-0017,

ACT-1-07-0028Operability Determination for CR 200603919, Fuel Oil Drain Line Leakage

A-3Attachment

A-4AttachmentSection 1R13CRs 20070202, 20070170, 20070197, 20070293, 20070703, 20070484, 200700886Problem Report

PR 93.0467.018.C.30.2, "Miscellaneous Plant Areas Ventilation Quarterly", Revision 101.5.22, "Risk Assessment Process," Revision 8

EOOS Risk Report and Scheduler's Evaluation for 1/16/07

3.M. 3-74, "Remote Telemetering Calibration," Revision 41.3.12, "Notification and Recall of Personnel," Revision 401.5.17, "Conduct of Maintenance," Revision 25

MR 07110043Section 1R15CRs 20070074, 20070108, 20070184, 20070629M282, Heating, Ventilation and Air Conditioning Temperature Control Diagrams, Revision E161.3.34.11, "Shift Operations Management System (eSOMS)

LCO Module," Revision 2

UFSAR 10.18, Equipment Area Cooling SystemUFSAR 8.4, Auxiliary Power Distribution SystemMR

071028963.M. 3-24.15, "Valve Stem Lubrication," Revision 6

EN -OP-104, "Operability Determinations," Revision

1ODMI Implementation Action Plan, Drywell Leakage, Revision 92.1.15, Daily Log Test #52, Revision 184Section 1R17

ER [[03118068, Ability to Automatically the Drywell Sumps for Sump Pumps P-301A, P-301B,P-305A and P-305B, Revision 0Field Sketch - Drywell Equipment Drain Sump Level Field Notes, dated October 7, 2003Drawing M232, Radwaste Collection System, Revision 30Schematic Diagram E96, Radwaste System, Sheet 1A, Revision 0Schematic Diagram E96, Radwaste System, Sheet 1, Revision 19Schematic Diagram E96, Radwaste System, Sheet 2, Revision 12Functional Drawing]]

SM 423, Liquid Radwaste System, Sheet 2, Revision E2Functional Drawing

SM "423, Liquid Radwaste System, Sheet 3, Revision E4Wiring Diagram M1P464-14, Reactor Water Cleanup & Recirculation Control Panel C904,Revision E22IE Bulletin 79-08, Events Relevant to Boiling Water Power Reactors Identified during Three MileIsland Incident, dated April 14, 19792.2.125.1, "Reset of Primary and [[system" contains a listed "[" character as part of the property label and has therefore been classified as invalid. Isolations (Group I,]]

II ,

III ,

IV , V,

VI andVII)," Revisions 16, 17, 18, 19

A-5AttachmentCR

200707000DRN -06-001626 through

DRN -06-016030 and 06-01869EN-LI-100,

ATT 9.1, Process Applicability Determination for

ER [[]]

03118068EN -

LI -100,

ATT 9.2, Impact Determination Questions for

ER [[]]

03118068EN -

LI -101,

ATT 9.1, 50.59 Reviews for

ER [[]]

03118068EN -

LI -110,

ATT 9.4, Commitment Change Evaluation Form for

ER [[]]

03118068EN -

LI -113,

ATT 9.1,

LBDCR Form for

ER 03118068

BEC o Letter #79-79, Response to

IE Bulletin 79-08, dated April 25, 1979

BEC o Letter #79-165, Supplementary Information to

IE Bulletin79-08 dated August 21, 1979General Electric Letter G-

HK-9-38, IE Bulletin 79-08, Events Relevant to Boiling Water PowerReactors Identified During Three Mile Island Incident Dated April 14, 1979, dated April

20, 1979Temporary Procedure

TP 03-034, "Alternate Drywell Equipment Sump Pumping Method,"Revision 3Section 1R193.M.3-61.5, "

EDG 2 Year Overhaul Preventive Maintenance," Revision

313.M. 3-61.2, "

EDG General and Preventive Maintenance Corrective Actions," Revision

293.M. 4-10, "Valve Maintenance," Revision 68.I.1.1, "Inservice Pump and Valve Testing Program," Revision 208.I.1, "Administration of Inservice Pump and Valve Testing," Revision 188.6.5.2, "Reactor Water Cleanup Valve Quarterly Operability," Revision 15

CRs 20070230, 20070233, 20070240, 20070241, 20070243, 20070247, 20070250, 20070252,20070254, 20070259, 20070260, 20070271, 20070278, 20070282, 20070289,

20070293, 20070294,

20070305MR s 04116408, 02119435, 02119532, 07101429, 07104403

TP 07-025, "Special Test for

EDG B Governor Adjustment or Replacement Postwork Testing,"Revision 0

CRs 20070703, 20070708, 20070714, 20070715, 20070718Section 1R228.5.5.9, "RCIC Simulated Automatic Actuation, Flow Rate, and Cold Quickstart Test,"Revision 168.9.1, "EDG and Associated Emergency Bus Surveillance," Revision 102 8.M.2-1.5.8.3, "Logic System Functional Test of System "A" Standby Gas Treatment Initiation,Reactor Building Isolation and Inboard Drywell Isolation Valves (Atmospheric Control

Valves)," Revision 288.5.1.1, "Core Spray System Operability - Pump Quarterly and Biennial Comprehensive FlowRate Tests and Valve Tests," Revision 438.A.18, "Core Spray System Integrity Surveillance," Revision 11

A-6Attachment8.I.1.1, "Inservice Pump and Valve Testing Program," Revision 20Updated Final Safety Analysis Section 9.4, Gaseous Radwaste System, Revision 217.3.36, "Offgas Sampling and Analysis," Revision 547.4.63, "Process Radiation Monitor Setpoints," Revision

43.M. 2-7.6, "

NUMAC [[]]

LOG Radiation Monitor Setpoint Change Procedure," Revision 10Windows Chemistry Data Management System (Win

CDMS ) DatabaseCRs 20070784, 20070074, 200700108, 20070703,

200707848.I. 1, "Admin of Inservice Pump and Valve Testing," Revision 18Section 1R23Temporary Alteration 06-1-064, Drill Through Disc of 29-

HO -3AER 06114160, Install Plexiglass Where Dampers Have Been Removed, Revision 02.2.45, "Screenhouse Heating and Ventilation System," Revision

188.C. 30.2, "Miscellaneous Plant Areas Ventilation Quarterly," Revision 11

UFSAR Table 10.9-1, Design Temperatures (Winter)UFSAR Table 10.9-2, Design Temperatures (Summer)Problem Report

PR 93.0467

UFSAR Change Request No. 2702Office Memorandum

FS&MC 93-108, Problem Report 93-0467.02 - Winterizing Intake Structure,dated November 23, 1993.Section 4

OA2CRs 20070193, 20070979, 20070086, 20070988, 20070996

Section

4OA 3Alarm Response Procedure

ARP -CP600R-B8, Process Rad Monitors 1705-18A,B (C19) 9.32,"Power Suppression Testing," Revision

8ODMI Implementation Action Plan, Fuel Defect, Revision 102.1.14, "Station Power Changes," Revision 92Emergency Operating Procedure

EOP -01, "Reactor Pressure Vessel Control," Revision 91.3.37, Scram Report 07-012.2.125.1, "Reset of Primary and Secondary Containment Isolations," Revision 18Power Maneuvering Plan

MAN.C 16-84, Revision 19Technical Specifications 3.6.C.1, 3.7.A.8

CRs 20070309, 20070312, 20070313, 20070322, 20070338, 20070344, 20070352, 20070353,20070354, 20070356, 20070366, 20070363, 20070364, 20070365, 20070367,

20070368, 20070390, 20070313, 20070377, 20070418, 20070419, 20070934,

20070938, 20070939, 20070943, 20070944, 20070947, 20070949, 20070952, 20070958

A-7AttachmentSection

4OA 7

ODCM Section 3.1.2 and 3.3.1, Revision 9WinCDMS32 (Windows Chemistry Data Management System)UFSAR 9.4, Gaseous Radwaste System, Revision 21Technical Specification 3.8.1, Main Condenser Offgas, Amendment

1773.M. 2-7.6, "

NUMAC Log Radiation Monitor Setpoint Change Procedure," Revision

8CR 200700784Apparent Cause Evaluation Report for

CR [[]]

200700784LIST [[]]

OF [[]]

ACRONY [[]]

MSADAMSAgencywide Documents Access and Management SystemCRcondition reportCScore sprayEDGemergency diesel generatorFdegrees Fahrenheitftfootgpmgallons per minuteHPCIhigh pressure coolant injectionIMCinspection manual chapterIRinspection reportkWkilowattkVkilovoltLCOlimiting condition of operationMRmaintenance requestNCVnon-cited violationNRCNuclear Regulatory CommissionOAother activitiesO&Moperations and maintenanceODCMoffsite dose calculation manualPARSPublicly Available RecordsPMTpost maintenance testRBCCWreactor building closed cooling waterRCICreactor core isolation coolingREO reasonable expectation for operabilityRFOrefueling outageRHRresidual heat removalRWCUreactor water cleanupSBGTstandby gas treatmentSBODGstation blackout diesel generator

A-8AttachmentSDPsignificant determination processSPARstandardized plant analysis riskSRAsenior reactor analystSSCsystem, structure or componentSSWsalt service waterTStechnical specificationsUFSARUpdated Final Safety Analysis Report