IR 05000445/2007005

IR 05000445-07-005 and 05000446-07-005, on 09/22/07 - 12/31/07, Comanche Peak, Units 1 and 2, Integrated Resident and Regional Report; Access Control to Radiologically Significant Areas, Identification & Resolution of Problems

ML080400011
Person / Time
Site:	Comanche Peak
Issue date:	02/08/2008
From:	Clay Johnson NRC/RGN-IV/DRP/RPB-A
To:	Blevins M Luminant Generation Co
References
IR-07-005
Download: ML080400011 (39)

Text

UNITED STATESNUCLEAR REGULATORY COMMISSION611 RYAN PLAZA DRIVE, SUITE 400ARLINGTON, TEXAS 76011-4005February 8, 2008Mike Blevins, Executive Vice President and Chief Nuclear OfficerLuminant Generation Company, LLCATTN: Regulatory AffairsComanche Peak Steam Electric StationP.O. Box 1002Glen Rose, TX 76043SUBJECT:COMANCHE PEAK STEAM ELECTRIC STATION - NRC INTEGRATEDINSPECTION REPORT 05000445/2007005 AND 05000446/2007005

Dear Mr. Blevins:

On December 31, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed aninspection at your Comanche Peak Steam Electric Station, Units 1 and 2, facility. The enclosedintegrated report documents the inspection findings that were discussed on January 2, 2008,with Tim Hope, Nuclear Licensing Manager, and members of your staff.This inspection examined activities conducted under your licenses as they relate to safety andcompliance with the Commission's rules and regulations, and with the conditions of yourlicenses. The inspectors reviewed selected procedures and records, observed activities, andinterviewed personnel.Based on the results of this inspection, two self-revealing findings of very low safety significance(Green) were identified. Both of these findings were determined to involve violations of NRCrequirements. Additionally, licensee-identified violations which were determined to be of verylow safety significance are listed in this report. However, because of the very low safetysignificance and because they are entered into your corrective action program, the NRC istreating these findings as noncited violations (NCVs) consistent with Section VI.A.1 of the NRCEnforcement Policy. If you contest any NCV in this report, you should provide a response within30 days of the date of this inspection report, with the basis for your denial, to the U.S. NuclearRegulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; withcopies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 611Ryan Plaza Drive, Suite 400, Arlington, Texas 76011-4005; the Director, Office of Enforcement,U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC ResidentInspector at the Comanche Peak Steam Electric Station.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 ins pection in theNRC Public Document Room or from the Publicly Available Records (PARS) component of Luminant Power-2-NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site athttp://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Claude E. Johnson, ChiefProject Branch ADivision of Reactor ProjectsDockets: 50-445 50-446Licenses: NPF-87 NPF-89

Enclosure:

NRC Inspection Report 05000445/2007005 and 05000446/2007005 w/attachment: Supplemental Information

REGION IVDockets:50-445, 50-446 Licenses:NPF-87, NPF-89Report:05000445/2007005 and 05000446/2007005Licensee:Luminant Generation Company, LLCFacility:Comanche Peak Steam El ectric Station, Units 1 and 2Location:FM-56, Glen Rose, TexasDates:September 22 through December 31, 2007Inspectors:D. Allen, Senior Resident InspectorA. Sanchez, Resident InspectorJ. Dixon, Senior Resident Inspector, South Texas ProjectS. Alferink, Reactor Inspector, Engineering Branch 2D. Stearns, Health PhysicistP. Elkmann, Emergency Preparedness InspectorApproved By:C. E. Johnson, Chief, Project Branch ADivision of Reactor ProjectsAttachment:Supplemental Information Enclosure-2-

SUMMARY OF FINDINGS

IR 05000445/2007005, 05000446/2007005; 09/22/2007-12/31/2007; Comanche Peak SteamElectric Station, Units 1 and 2. Integrated Resident and Regional Report; Access Control toRadiologically Significant Areas, Identification & Resolution of Problems.This report covered a 3-month period of inspection by three resident inspectors, an EngineeringBranch reactor inspector, an emergency preparedness inspector, and an occupational radiationsafety inspection by a health physicist. Two Green self-revealing findings, both of which werenoncited violations, were identified. Three licensee identified violations are also documented inthis report. The significance of most findings is indicated by their color (Green, White, Yellow,or Red) using Inspection Manual Chapter 0609 "Significance Determination Process." Findingsfor which the Significance Determination Process does not apply may be Green or be assigneda severity level after NRC management review. The NRC's program for overseeing the safeoperation of commercial nuclear power reactors is described in NUREG-1649, "ReactorOversight Process," Revision 3, dated July 2000.A.

Cornerstone: Initiating Events

• Green: A Green self-revealing noncited violation of Technical Specification5.4.1.a was identified for the failure to use a procedure appropriate to thecircumstances when performing maintenance on safety-related equipment. Specifically, on October 22, 2006, the licensee used a procedure not appropriateto the circumstances when making adjustments to the Exhaust Pilot Valve 2-HV-2452-1-PR3 on the Main Steam Line 2-04 to Auxiliary Feedwater Pump TurbineSteam Supply Valve 2-HV-2452-1. The adjustments to the exhaust pilot valveeventually led to three inadvertent operations of the turbine driven auxiliaryfeedwater pump on March 12, 2007. The licensee entered the finding into theircorrective action program. One corrective action included adding additionalinformation and guidance to the procedures.This issue was determined to be more than minor because it is similar toExample b of Section 4, "Insignificant Procedural Errors," in ManualChapter 0612, Appendix E, "Examples of Minor Issues." Specifically, this issue ismore than minor because it led to a plant transient that resulted in a reduction inreactor power. Additionally, this issue is associated with the Initiating Eventscornerstone attribute of human performance and affected the cornerstoneobjective to limit the likelihood of those events that upset plant stability andchallenge critical safety functions during power operations. This finding wasdetermined to be of very low safety significance because the finding did notcontribute to both the likelihood of a reactor trip and the likelihood that mitigationequipment or functions would not be available. The cause of the finding isrelated to the cross-cutting aspect of Human Performance in that the licenseefailed to use a systematic decision making process to determine unintendedconsequences that would occur in decreasing the stroke time of the exhaust pilotvalve (H.1.(a)) (Section 4OA2).

Enclosure-3-

Cornerstone: Occupational Radiation Safety

Green.

The inspector reviewed a self-revealing noncited violation of10 CFR 20.1501(a) for failure to conduct a radiological survey. Specifically, onApril 16, 2007, a worker's electronic dosimeter alarmed when the individualattempted to move a bag containing a small vacuum cleaner from a postedcontaminated and radiation area. The bag of materials had not been surveyedfor radiation levels and, therefore, had not been labeled to indicate the potentialhazard. The bag was subsequently surveyed and found to have radiation levelsof 600 millirem per hour on contact and 150 millirem per hour at 30 c entimetersfrom the surface. Corrective actions include counseling of personnel, evaluationof possible organizational changes, and generation of a training request toinclude this event in future training.

The failure to conduct a radiological survey is a performance deficiency. Thisfinding is greater than minor because it is associated with the OccupationalRadiation Safety Program and Process attribute and affected the cornerstoneobjective, which is to ensure adequate protection of worker health and safetyfrom exposure to radiation. The failure to perform the radiation survey led to aworker receiving unintended and additional exposure. Using the OccupationalRadiation Safety Significance Determination Process, the inspector determinedthat the finding was of very low safety significance because it did not involve: (1) as low as is reasonably achievable planning and controls, (2) anoverexposure, (3) a substantial potential for overexposure, or (4) an impairedability to assess dose. In addition, this finding has a cross-cutting componentassociated with human performance and work coordination because the licenseefailed to keep workers apprised of work status and plant conditions that mayaffect work activities prior to removing contaminated items from the reactorcontainment building (H.3(b)) (Section 2OS1).

B.Licensee-Identified Violations

Violations of very low safety significance which were identified by the licensee havebeen reviewed by the inspectors. Corrective actions taken or planned by the licenseehave been entered into their corrective action program. These violations and correctiveactions are listed in Section 4OA7 of this report.

Enclosure-4-

REPORT DETAILS

Summary of Plant StatusComanche Peak Steam Electric Station (CPSES) Unit 1 began the inspection period operatingat 100 percent power.

On December 13, while performing the monthly control rodrepositioning, Control Rod J13 (in Shutdown Bank B) dropped to the full in position. The reactorpower was subsequently reduced to less than 50 percent. Upon recovery of the control rod,reactor power was restored to 100 percent on December 15, 2007.CPSES Unit 2 operated at 100 percent power for the duration of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity1R04Equipment Alignment

.1 Partial System Walkdown

a. Inspection Scope

The inspectors:

(1) walked down portions of the below listed risk important systems andreviewed plant procedures and documents to verify that critical portions of the selectedsystems were correctly aligned and

(2) compared deficiencies identified during thewalkdown to the licensee's corrective action program to ensure problems were beingidentified and corrected.

• Unit 1 Residual Heat Removal (RHR) System 1-01 in accordance with SystemOperating Procedures Manual (SOP) SOP-1-02A, "Residual Heat RemovalSystem," Revision 15, while RHR System 1-02 (Train B) was inoperable forscheduled maintenance and surveillanc e testing on November 1, 2007The inspectors completed one sample.

b. Findings

No findings of significance were identified.

.2 Detailed Semiannual System Walkdown

a. Inspection Scope

The inspectors conducted a detailed semiannual inspection of the control room airconditioning system, to verify the functional capability of the system. The inspectorsreferenced and used the following documents to verify the proper system alignment,electrical power supply, and setpoints:*SOP-802, "Control Room Ventilation System," Revision 11

-5-*Drawing M1-0304, "Ventilation Control Room Air Conditioning," Revision CP-34*Operations Department Administration Manual ODA-308, "LCO TrackingProgram," Revision 11*Design Basis Document DBD-ME-304, "Control Room Air Conditioning System,"Revision 17*Design Basis Document DBD-ME-003, "Control Room Habitability," Revision 10The inspectors also reviewed recent corrective action documents, system health reports,outstanding work requests, and design issues to determine if any of these items impactthe system's ability to operat e as designed or i ndicated a degradation in capability. Acomplete field walkdown was performed by the inspectors during the week of December 17, 2007.The inspectors completed one sample.

b. Findings

No findings of significance were identified.

1R05 Fire Protection (71111.05) Fire Area Tours

a. Inspection Scope

The inspectors walked down the plant areas listed below to assess the materialcondition of active and passive fire protection features and their operational lineup andreadiness. The inspectors:

(1) verified that transient combustibles and hot workactivities were controlled in accordance with plant procedures;

(2) observed thecondition of fire detection devices to verify they remained functional;

(3) observed firesuppression systems to verify they remained functional;

(4) verified that fireextinguishers and hose stations were provided at their designated locations and thatthey were in a satisfactory condition;

(5) verified that passive fire protection features(electrical raceway barriers, fire doors, fire dampers, steel fire proofing, penetrationseals, and oil collection systems) were in a satisfactory material condition;

(6) verifiedthat adequate compensatory measures were established for degraded or inoperable fireprotection features; and

(7) reviewed the corrective action program to determine if thelicensee identified and corrected fire protection problems.

• Fire Zone SG010 - Unit 1 Train A Emergency Diesel Generator (EDG) rooms 84,99B and 99D on 810 foot and 844 foot elevations on October 9, 2007*Fire Zone AA21F - Auxiliary Building 852 foot elevation on Nove mber 4, 2007

-6-*Fire Zone 2SB4 - Unit 2 Safeguards Building 790 foot elevation onNovember 5, 2007*Fire Zones EA54, EA57 through EA61 - Electrical and Control Building 792 footelevation, Units 1 and 2 Train C inverter and battery rooms on November 6, 2007

• Fire Zone 2SI012 - Unit 2 Train B EDG rooms 2-85, 2-99A, and 2-99C onNovember 6, 2007The inspectors completed five samples.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification (71111.11Q)

a. Inspection Scope

On October 25, 2007, the inspectors observed examination simulator scenarios with twodifferent operating crews. The scenario began at 100 percent reactor power. Thefollowing events then took place:

(1) a heater drain pump failure causing a runback to800 MWe;

(2) an unisolable condenser vacuum leak that caused the operators tomanually trip the reactor;

(3) a main steam leak from Steam Generator 1-4 outside ofcontainment; and

(4) an automatic Phase A containment isolation failure. The scenariowas terminated following the crew demonstrating its ability to isolate the faulted steamgenerator. An Alert emergency classification was declared and the notification wasmade in a timely manner. Simulator observations included formality and clarity of communications, groupdynamics, the conduct of operations, procedure usage, command and control, andactivities associated with the emergency plan. The inspectors also verified thatevaluators and operators were identifying crew performance deficiencies as applicable. The inspectors also reviewed the quality of the examination scenario for qualitative andquantitative attributes according to Appendix A of this inspection procedure.The inspectors completed one sample.

b. Findings

No findings of significance were identified.

-7-1R12Maintenance Effectiveness (71111.12) Routine Maintenance Effectiveness Inspection

a. Inspection Scope

The inspectors reviewed the two maintenance activities listed below to:

(1) verify theappropriate handling of structure, system, and component (SSC) performance orcondition problems;

(2) verify the appropriate handling of degraded SSC functionalperformance;

(3) evaluate the role of work practices and common cause problems; and(4) evaluate the handling of SSC issues reviewed under the requirements of theMaintenance Rule, 10 CFR Part 50, Appendix B, and the Technical Specifications (TS).*Unit 2 Component Cooling Water Heat Exchanger 2-02 emergent heatexchanger cleaning activities stemming from macro-fouling that causedunexpected unavailability, documented in Smart Form (SMF) SMF-2007-2444-00*Unit 1 and 2 atmospheric relief valve isolation valves' remote operatormodification that resulted in a number of hours of unavailability for the system The inspectors completed two samples.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13)

a. Inspection Scope

The inspectors reviewed the five below listed assessment activities to verify:

(1) performance of risk assessments when required by 10 CFR 50.65(a)(4) and siteprocedures prior to changes in plant configuration for maintenance activities and plantoperations;

(2) the accuracy, adequacy, and completeness of the information consideredin the risk assessment;

(3) that the licensee recognizes, and/or enters as applicable, theappropriate risk category according to the risk assessment results and site procedures;and

(4) the licensee identified and corrected problems related to maintenance riskassessments.*Emergent work for troubleshooting Turbine Driven Auxiliary Feedwater (TDAFW)Steam Admission Valve 2-HV-2452-2, in conjunction with the rescheduledremote shutdown panel operability test of the auxiliary feedwater (AFW) system,reviewed on October 11, 2007*Scheduled and emergent work activities for the week of October 7 through 13,2007, including scheduled outage of offsite power via Transformer XST1, androutine surveillances of EDG 1-01 and TDAFW Pump 2-01, reviewed onOctober 15, 2007

-8-*Scheduled and emergent work activities for the week of October 14 through 20,2007, including scheduled outage of the 138 kV line from the DeCordova stationand routine surveillances of EDG 2-01 and TDAFW Pump 1-01, reviewed onOctober 15, 2007 *Scheduled EDG 2-02 surveillance testing with the 138 kV DeCordova power linede-energized for line and tower modification (at the DeCordova substation),which required the safety-related buses to be realigned to their alternate powersource, Transformer XST2 (supplied from the 345 kV switchyard), reviewed onOctober 31, 2007*Scheduled 345 kV Transformer XST2 maintenance outage, scheduledsurveillance run of EDG 2-02, and the emergent corrective maintenance onEDG 1-02 to replace Fuel Pump 6L, reviewed November 26-28, 2007The inspectors completed five samples.

b. Findings

No findings of significance were identified.

1R15 Operability Ev

aluations (71111.15)

a. Inspection Scope

The inspectors:

(1) reviewed plant status documents such as operator shift logs,emergent work documentation, deferred modifications, and standing orders to determineif an operability evaluation was warranted for degraded components;

(2) referred to theUFSAR and design basis documents to review the technical adequacy of licenseeoperability evaluations;

(3) evaluated compensatory measures associated withoperability evaluations;

(4) determined degraded component impact on any TechnicalSpecifications;

(5) used the significance determination process (SDP) to evaluate therisk significance of degraded or inoperable equipment; and

(6) verified that the licenseehas identified and implemented appropriate corrective actions associated with degradedcomponents. The inspectors interviewed appropriate licensee personnel to provideclarity to operability ev aluations, as necessary. Specific operability evaluations reviewedare listed below:*SMF-2007-3098-00, documents the operability determination for piping insulationnot included in the Combustible Loading Calculation 0210-063-0002 for Unit 1Train A containment spray room, reviewed on November 26, 2007*SMF-2007-2218-00, documenting incorrect information used in the calculationsfor the Units 1 and 2 refueling water storage tank setpoints and the minimumcontainment flood levels, and the corrective actions taken, reviewed onDecember 4, 2007

-9-*SMF-2007-3107-00, documenting Main Steam Line 1-01 before main steamisolation valve drip pot isolation Valve 1-HV-2409 closing stroke time to be in thealert range, reviewed on December 5, 2007*SMF-2007-2909-00, documents the operability determination of the Unit 2TDAFW pump following the discovery of a leak in the 2-HV-2452-2 steamadmission valve, reviewed on December 11-12, 2007The inspectors completed four samples.

b. Findings

No findings of significance were identified.

1R19 Post-maintenance Testing (71111.19)

a. Inspection Scope

The inspectors selected the below listed post-maintenance test activity of a risk-significant component. The inspectors:

(1) reviewed the applicable licensing basisand/or design basis documents to determine the safety functions,

(2) evaluated thesafety functions that may have been affected by the maintenance activity, and(3) reviewed the test procedure to ensure it adequately tested the safety function thatmay have been affected. The inspectors either witnessed or reviewed test data to verifythat acceptance criteria were met, plant impacts were evaluated, test equipment wascalibrated, procedures were followed, jumpers were properly controlled, the test dataresults were complete and accurate, the test equipment was removed, the system wasproperly realigned, and deficiencies during testing were documented. The inspectorsalso reviewed the UFSAR to determine if the licensee identified and corrected problemsrelated to post-maintenance testing.*Unit 1 RHR Valve 1-8812A, Refueling Water Storage Tank to RHR Pump 1-01suction valve, in accordance with Operations Testing Manual (OPT) OPT-512A,"RHR and SI Subsystem Valve Test," Revision 9, following a major inspection ofthe valve performed on October 11, 2007The inspectors completed one sample.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing (71111.22)

a. Inspection Scope

The inspectors evaluated the adequacy of periodic testing of important nuclear plantequipment, including aspects such as preconditioning, the impact of testing during plantoperations, and the adequacy of acceptance criteria. Other aspects evaluated included

-10-test frequency and test equipment accuracy, range, and calibration; procedureadherence; record keeping; the restoration of standby equipment; test failureevaluations; system alarm and annunciator functionality; and the effectiveness of the licensee's problem identification and correction program. The following surveillance testactivities were observed and/or reviewed by the inspectors:*Unit 2 remote shutdown panel surveillance test in accordance with OPT-216B,"Remote Shutdown Operability Test," Revision 9, observed on Oct ober 11, 2007*Unit 1 RHR Pump 1-01 (inservice test) surveillance in accordance with OPT-203A, "Residual Heat Removal System," Revision 15, observed onOctober 11, 2007The inspectors completed two samples.

b. Findings

No findings of significance were identified.

===Cornerstone: Emergency Preparedness1EP4Emergency Action Level and Emergency Plan Changes (71114.04)

a. Inspection Scope

The inspector verified that Inspection Procedure 71114.04 was not documented in aninspection report during the period January through December 2007 because,

(1) thelicensee did not submit changes to their Emergency Plan or emergency planimplementing procedures that required regulatory review, or

(2) all changes to thelicensee's emergency plan and implementing procedures implemented during theinspection period were determined to be of minor significance, not requiringdocumentation according to the criteria of Inspection Procedure 71114.04. The inspector completed one sample.

b. Findings

No findings of significance were identified.1EP6Drill Evaluati===

on (71114.06 and 71114.07)

a. Inspection Scope

For the two below listed drill and simulator-based training ev olutions c ontributing todrill/exercise performance (DEP) and emergency response organization drill participation(ERO) performance indicators, the inspectors:

(1) observed the training evolution toidentify any weaknesses and deficiencies in classification, notification, and protectiveaction recommendations development activities;

(2) compared the identified weaknessesand deficiencies against licensee identified findings to determine whether the licensee is

-11-properly identifying failures; and

(3) determined whether licensee performance is inaccordance with the guidance of the Nuclear Energy Institute (NEI) 99-02, "RegulatoryAssessment Performance Indicator Guideline," Revision 5, acceptance criteria.*October 17, 2007, Force-on-Force exercise, Day Two

• November 8, 2007, emergency preparedness exercise, observed from thesimulator and the emergency operations facility, with the occurrence of anearthquake and subsequent damage to the spent fuel pool, resulting in thedeclaration of a General Emergency.The inspectors completed two samples.

b. Findings

No findings of significance were identified.2.RADIATION SAFETYCornerstone: Occupational Radiation Safety 2OS1Access Control To Radiologically Significant Areas (71121.01)

a. Inspection Scope

This area was inspected to assess the licensee's performance in implementing physicaland administrative controls for airborne radioactivity areas, radiation areas, highradiation areas, and worker adherence to these controls. The inspector used therequirements in 10 CFR Part 20, the Technical Specifications, and the licensee'sprocedures required by technical specifications as criteria for determining compliance. During the inspection, the inspector interviewed the radiation protection manager,radiation protection supervisors, and radiation workers. The inspector performedindependent radiation dose rate measurements and reviewed the following items:*Performance indicator events and associated documentation packages reportedby the licensee in the Occupational Radiation Safety Cornerstone*Self-assessments, audits, licensee event reports, and special reports related tothe access control program since the last inspection*Adequacy of the licensee's internal dose assessment for any actual internalexposure greater than 50 millirem committed effective dose equivalent *Corrective action documents related to access controls

• Radiation work permit briefings and worker instructions

• Changes in licensee procedural controls of high dose rate - high radiation areasand very high radiation areas

-12-*Posting and locking of entrances to all accessible high dose rate - high radiationareas and very high radiation areas*Radiation worker and radiation protection technician performance with respect toradiation protection work requirements The inspector completed 12 of the required 21 samples.

b. Findings

Introduction.

The inspector reviewed a self-revealing noncited violation of10 CFR 20.1501(a) for the failure to evaluate the actual radiological hazards beforeremoving a bag of items from a posted area in containment.

Description.

On April 16, 2007, a worker's electronic dosimeter alarmed when theindividual attempted to move a bag containing a small vacuum cleaner from a postedarea inside containment to the control point located at the entrance to containment. Thebag of materials had not been surveyed for radiation levels and therefore had not beenlabeled to indicate the potential hazard. After performance of an integrated leak rate test a small portable vacuum used to cleanup broken glass was left in a posted contaminated area near the refuel cavity. Adiscussion was held between the decontamination technicians and the radiationprotection technician concerning the cleanup of the broken glass, but the radiologicalstatus of the vacuum was not discussed. The radiation protection technician intended tosurvey the vacuum, but other activities distracted the technician from evaluating the area. After shift turnover, containment cleanup started, including removal of material frominside the posted contaminated area on the 860' elevation near the reactor cavity. Arrangements were made with radiation protection and maintenance services to removethe items from the contaminated area without a survey, and to perform the survey at thecontainment exit area on the 832' elevation. The vacuum was placed in a bag at thecontamination area boundary. During transport of the bag to the control point on the832' elevation, a maintenance service person received an unanticipated dose rate alarmfrom the bag containing the vacuum. Surveys performed on the vacuum indicated 600 millirem per hour on contact and 150 millir em per hour at 30 centimeters. Ascorrective action, the licensee counseled personnel who authorized the tagging ofmaterial at a location other than at the contamination area, generated actions to considerorganizational changes, and generated a training request to include this event in futuretraining.Analysis. The failure to perform a radiological survey is a performance deficiency. Thisfinding is greater than minor because it is associated with the Occupational RadiationSafety Program and Process attribute and affected the cornerstone objective, which is toensure adequate protection of worker health and safety from exposure to radiation. Thefailure to perform the radiation survey led to a worker receiving unintended and additionalexposure. Using the occupational radiation safety significance determination process,the inspector determined that the finding was of very low safety significance because it

-13-did not involve:

(1) as low as is reasonably achievable (ALARA) planning and controls,(2) an overexposure,

(3) a substantial potential for overexposure, or

(4) an impairedability to assess dose. In addition, this finding had a cr osscutting component associatedhuman performance and work coordination because the licensee failed to keep workersapprised of work status and plant conditions that may affect work activities prior toremoving contaminated items from the reactor containment building (H.3(b)).Enforcement. Part 20.1501(a) of Title 10 of the Code of Federal Regulations states thateach licensee shall make, or cause to be made, surveys that:

(1) may be necessary forthe licensee to comply with the regulations in this part; and

(2) are reasonable under thecircumstances to evaluate the magnitude and extent of radiation levels, concentrations orquantities of radioactive material, and the potential radiological hazards. Part 20.1003 ofTitle 10 of the Code of Federal Regulations defines survey as an evaluation of theradiological conditions and potential hazards incident to the production, use, transfer,release, disposal, or presence of radioactive material or other sources of radiation. Part 20.1201 requires the licensee to control the occupational dose to individual adults. Contrary to these requirements, radiation protection technicians failed to survey a bag ofcontaminated material prior to workers moving the bag from a posted contaminated area. Because the failure to perform a radiological survey is of very low safety significance andhas been entered into the licensee's corrective action program (Smart Form 2007-1337-00), this violation is being treated as an NCV, consistent with Section VI.A.1 of theNRC Enforcement Policy: (NCV 05000445;446/2007005-01, "Failure To EvaluateRadiological Conditions").2OS2ALARA Planning and Controls (71121.02)

a. Inspection Scope

The inspector assessed licensee performance with respect to maintaining individual andcollective radiation exposures ALARA. The inspector used the requirements in 10 CFRPart 20 and the licensee's procedures required by technical specifications as criteria fordetermining compliance. The inspector interviewed licensee personnel and reviewed:*Current 3-year rolling average collective exposure

• Site-specific trends in collective exposures, plant historical data, and source-termmeasurements*Site-specific ALARA procedures

• Five work activities of highest exposure significance completed during the lastoutage*ALARA work activity evaluations, exposure estimates, and exposure mitigationrequirements*Intended versus actual work activity doses and the reasons for anyinconsistencies

-14-*Person-hour estimates provided by maintenance planning and other groups to theradiation protection group with the actual work activity time requirements *Post-job (work activity) reviews

• Assumptions and basis for the current annual collective exposure estimate, themethodology for estimating work activity exposures, the intended dose outcome,and the accuracy of dose rate and man-hour estimates*Method for adjusting exposure estimates, or re-planning work, when unexpectedchanges in scope or emergent work were encountered*Records detailing the historical trends and current status of tracked plant sourceterms and contingency plans for expected changes in the source term due tochanges in plant fuel performance issues or changes in plant primary chemistry *Radiation worker and radiation protection technician performance during workactivities in radiation areas, airborne radioactivity areas, or high radiation areas *Self-assessments, audits, and special reports related to the ALARA programsince the last inspection*Corrective action documents related to the ALARA program and follow-upactivities, such as initial problem identification, characterization, and tracking *Effectiveness of self-assessment activities with respect to identifying andaddressing repetitive deficiencies or significant individual deficiencies The inspector completed 12 of the required 15 samples and 3 of the optional samples.

b. Findings

No findings of significance were identified.4.OTHER ACTIVITIES

4OA1 Performance Indicator (PI) Verification

.1 Mitigating Systems Cornerstone

a. Inspection Scope

The inspector reviewed a sample of PI data submitted by the licensee regarding themitigating system cornerstone to verify that the licensee's data was reported inaccordance with the requirements of NEI 99-02, "Regulatory Assessment PerformanceIndicator Guideline," Revision 5. Reactor operator logs, limiting condition for operationaction requirement logs, EVAL-2006-2236-01, EVAL-2006-2750-01, SMF-2006-2750-00,

-15-EVAL-2006-3636-01, EVAL-2007-1172-01, EVAL-2007-2208-01, EVAL-2007-2846-01,and licensee event reports submitted between April 2006 and September 2007, werereviewed for both Units 1 and 2 to identify for the following PI:*Units 1 and 2 Safety System Functional Failures The inspectors completed two samples in this cornerstone.

b. Findings

No findings of significance were identified.

.2 Occupational Radiation Safety Cornerstone

a. Inspection Scope

Occupational Exposure Control Effectiveness The inspector reviewed licensee documents from April 1 through September 30, 2007. The review included corrective action documentation that identified occurrences in lockedhigh radiation areas (as defined in the licensee's technical specifications), very highradiation areas (as defined in 10 CFR 20.1003), and unplanned personnel exposures (asdefined in NEI 99-02, "Regulatory Assessment Indicator Guideline," Revision 5). Additional records reviewed included ALARA records and whole body counts of selectedindividual exposures. The inspector interviewed licensee personnel that wereaccountable for collecting and evaluating the performance indicator data. In addition, theinspector toured plant areas to verify that high radiation, locked high radiation, and veryhigh radiation areas were properly controlled. Performance indicator definitions andguidance contained in NEI 99-02, Revision 5, were used to verify the basis in reportingfor each data element.The inspector completed the required sample

(1) in this cornerstone.

b. Findings

No findings of significance were identified.

.3 Public Radiation Safety Cornerstone

a. Inspection Scope

Radiological Effluent Technical Specification/Offsite Dose Calculation Manual Radiological Effluent Occurrences

-16-The inspector reviewed licensee documents from April 1 through September 30, 2007. Licensee records reviewed included corrective action documentation that identifiedoccurrences for liquid or gaseous effluent releases that exceeded performance indicatorthresholds and those reported to the NRC. The inspector interviewed licensee personnelthat were accountable for collecting and evaluating the performance indicator data. Performance indicator definitions and guidance contained in NEI 99-02, Revision 5, wereused to verify the basis in reporting for each data element.The inspector completed the required sample

(1) in this cornerstone.

b. Findings

No findings of significance were identified.4OA2Identification and Resolution of Problems (71152)

.1 Routine Review of Identification and Resolution of Problems

a. Inspection Scope

The inspectors performed a daily screening of items entered into CPSES's correctiveaction program. This assessment was accomplished by reviewing SMFs and event trendreports and attending daily operational meetings. The inspectors:

(1) verified thatequipment, human performance, and program issues were being identified by CPSES atan appropriate threshold and that the issues were entered into the corrective actionprogram;

(2) verified that corrective actions were commensurate with the significance ofthe issue; and

(3) identified conditions that might warrant additional follow-up throughother baseline inspection procedures.

b. Findings

No findings of significance were identified.

.2 Semiannual Trend Review

a. Inspection Scope

On December 21, 2007, the inspectors completed a semiannual review of licenseeinternal documents, reports, and performance indicators to identify trends that mightindicate the existence of more safety significant issues. The inspectors reviewed thefollowing types of documents:Corrective Action Documents (Smart Forms)System Health ReportsPlanned Maintenance Work Week CritiquesCPSES Nuclear Overview Department Evaluation Reports (Audits)

-17-Human Performance Program Health Indicators PackageCorrective Action Program Health reportStation Reliability IssuesDegraded conditions evaluated in accordance with Generic Letter 91-18CPSES Self-Assessment Reports

b. Findings and Observations

No findings of significance were identified. The inspectors did note that the licensee wasimplementing a new process for capturing and displaying equipment performanceparameters for the purpose of trending. The inspector's initial observation was that thisshould assist operations and engineering in reviewing data and identifying trends in amore timely manner. The inspector did not identify any additional trends.

.3 Selected Issue Follow-Up Inspection - Cumulative Effects of Operator Workarounds

a. Inspection Scope

The inspectors reviewed the cumulative effects of the operator workarounds to determine:

(1) the reliability, availability, and potential for misoperation of a system;

(2) if multiple mitigating systems could be affected;

(3) the ability of operators to respond in acorrect and timely manner to plant transients and accidents; and

(4) if the licensee hasidentified and implemented appropriate corrective actions associated with operatorworkarounds.

b. Findings

No findings of significance were identified.

.4 Selected Issue Follow-Up Inspection - Review of Inadvertent Actuations of Unit 2 TurbineDriven Auxiliary Feedwater Pump

a. Inspection Scope

The inspectors reviewed the circumstances, events, and issues surrounding threeinadvertent operations of the Unit 2 TDAFW pump on March 12, 2007. These actuationsled to a 60-day notification to the NRC Operations Center. This issue was selected forreview due to the risk significance of the TDAFW pump, and the event resulting in areduction in power and entry into a TS Limited Condition for Operation with a shutdownaction requirement.As part of the review, the inspectors reviewed procedures, work orders, and smart formsassociated with this event. The inspectors reviewed the root cause analysis to assessthe detail of the review, adequacy of the root cause, and proposed corrective actions.

-18-The licensee's investigation determined that the root cause of the actuations was themaintenance stroke timing restrictions. The inspectors discussed the events and analysiswith system engineering and the root cause analyst.The inspectors completed two samples of selected issue follow-up.

b. Findings and Observations

Introduction:

A Green self-revealing noncited violation of TS 5.4.1.a was identified for thefailure to use a procedure appropriate to the circumstances when performingmaintenance on safety-related equipment. Specifically, the licensee used a procedurenot appropriate to the circumstances when making adjustments to the Exhaust PilotValve 2-HV-2452-1-PR3 on the Main Steam Line 2-04 to Auxiliary Feedwater PumpTurbine Steam Supply Valve 2-HV-2452-1. The adjustments to the exhaust pilot valveeventually led to three inadvertent operations of the TDAFW pump.Description: On October 8, 2006, the licensee replaced the Exhaust PilotValve 2-HV-2452-1-PR3 and rebuilt the pressure regulator on Valve 2-HV-2452-1 as ascheduled preventive maintenance activity. These activities were performed under WorkOrder 03-05-341936-01. The as-found actuation pressure for the old exhaust pilot valveand the as-left actuation pressure for the new exhaust pilot valve were recorded as27.2 psig. The licensee did not perform post work testing of valve 2-HV-2452-1 at thattime.Around October 15, 2006, while in Mode 5, the licensee performed post work testing of Valve 2-HV-2452-1 using procedure OPT-206B, "AFW System," Revision 18,Section 8.2.3. Valve 2-HV-2452-1 failed to meet the post work test acceptance criteria byopening too slowly. On October 22, 2006, while in Mode 4, the licensee again performedpost work testing on Valve 2-HV-2452-1 using procedure OPT-206B. Again, the valvefailed to meet the post work test acceptance criteria by opening too slowly, and it becamea Mode 3 restraint to leaving the outage.The licensee assembled a team to resolve the slow opening of Valve 2-HV-2452-1. Theteam determined that reducing the actuation pressure of the exhaust pilot valve wouldallow the Valve 2-HV-2452-1 air actuator to exhaust faster, which would reduce theamount of time the valve took to open. The valve team worked under WorkOrder 03-05-341936-01 and rotated the exhaust pilot valve adjustment screw to decreasethe actuation pressure. Later testing determined that this adjustment decreased theactuation pressure to approximately 0 psig.Work Order 03-05-341936-01 included Data Sheet MDA-1105-3, "TSS Valve DataSheet," which contained information and instructions for Valve 2-HV-2452-1. The datasheet specifically noted that "minor adjustments of PR2 [throttle valve] needle valve maybe necessary in order to satisfy valve stroke time per OPT-206B." The data sheet,however, did not discuss adjustments to PR3 (exhaust pilot valve), nor did it provide clearguidance on acceptable limits for the actuation pressure for the exhaust pilot valve.Due to the small actuation pressure setpoints, the exhaust pilot valve eventually startedventing to atmosphere. On March 12, 2007, an operator heard air venting from the

-19-exhaust pilot valve and placed his hand close to the valve. This action restricted theventing air flow and increased the back pressure sensed by one of the valve's ports. Since the actuation pressure was set at such a low value, this small amount of backpressure was sufficient to actuate the valve. The pilot valve swapped, exhausting thediaphragm. Consequently, Valve 2-HV-2452-1 opened and the TDAFW pump startedand reached full flow. Operations personnel responded by manually running back themain turbine to 1100 MW to ensure reactor power remained less than 100%. Approximately 15 minutes later, Operations personnel secured the TDAFW pump, closedValve 2-HV-2452-1, and placed the associated Hand switch 2-HS-2452-1 in auto. ThePrompt Team was requested to determine the cause of the valve failure.While examining the valve actuator and tubing joints, a Prompt Team technician came inclose proximity to Valve 2-HV-2452-1 and restricted the venting air flow, increasing theback pressure sensed by one of the exhaust pilot valve's ports. Again, this small amountof back pressure caused the exhaust pilot valve to swap ports, exhausting the diaphragm. Consequently, Valve 2-HV-2452-1 opened and the TDAFW pump reached full flow asecond time.The licensee then closed the upstream isolation valve, Valve 2-MS-0128. While in theprocess of closing this valve, the operator heard air venting and placed his hand underthe exhaust pilot valve, increasing the back pressure sensed by one of the exhaust pilotvalve's ports. For a third time, this small amount of back pressure caused the exhaustpilot valve to swap ports, exhausting the diaphragm. Consequently, Valve 2-HV-2452-1opened a third time. This time, the TDAFW pump reached a partial flow because theupstream isolation valve was partially closed.Operations personnel initiated a clearance to prevent further inadvertent actuations of theTDAFW pump and a work order to determine the valve failure mechanism. OnMarch 13, 2007, the licensee declared the TDAFW pump operable. The licensee enteredthe inadvertent operations of the TDAFW pump into their corrective action program forresolution. The licensee performed several corrective actions for this finding, includingadding additional information and guidance to the MDA-1105 data sheet.Analysis: The adjustment of the exhaust pilot valve actuation pressure outside of theacceptable performance band was the performance deficiency. This issue wasdetermined to be more than minor because it is similar to Example b of Section 4,"Insignificant Procedural Errors," in Manual Chapter 0612, Appendix E, "Examples ofMinor Issues." Specifically, this issue is more than minor because it led to a planttransient that resulted in a reduction in reactor power. Additionally, this issue isassociated with the Initiating Events cornerstone attribute of human performance andaffected the cornerstone objective to limit the likelihood of those events that upset plantstability and challenge critical safety functions during pow er operations. This finding wasdetermined to be of very low safety significance because the finding did not contribute toboth the likelihood of a reactor trip and the likelihood that mitigation equipment orfunctions would not be available. The cause of the finding is related to the cross-cuttingaspect of Human Performance in that the licensee failed to use a systematic decisionmaking process to determine unintended consequences that would occur in decreasingthe stroke time of the exhaust pilot valve (H.1.(a)).

-20-Enforcement: Technical Specification 5.4.1.a states, in part, that "Written procedures shall be established, implemented, and maintained covering ...The applic able proceduresrecommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978." Regulatory Guide 1.33, Appendix A, Item 9 states, in part, that "Maintenance that canaffect the performance of safety-related equipment should be properly preplanned andperformed in accordance with written procedures, documented instructions, or drawingsappropriate to the circumstances." Contrary to the above, on October 22, 2006, thelicensee failed to use a procedure appropriate to the circumstances when performingmaintenance on safety-related equipment. Specifically, the licensee used a procedurenot appropriate to the circumstances when making adjustments to the Exhaust PilotValve 2-HV-2452-1-PR3 on the Main Steam Line 2-04 to Auxiliary Feedwater PumpTurbine Steam Supply Valve 2-HV-2452-1. The adjustments to the exhaust pilot valveeventually led to three inadvertent operations of the TDAFW pump. Because this findingis of very low safety significance and because it was entered into the licensee's correctiveaction program as Smart Form SMF-2007-0903-00, this violation is being treated as anon-cited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV05000446/2007005-02, Failure to Use a Procedure Appropriate to the Circumstances Ledto Inadvertent Actuations of the Turbine Driven Auxiliary Feedwater Pump.

.1 Unit 1 Dropped Control Rod

a. Inspection Scope

During the performance of monthly rod repositioning on December 13, 2007, the Unit 1 Control Rod J13 in Shutdown Bank B dropped to the full-in position with the remainingShutdown Bank B rods in the full out position. The inspectors responded to the Unit 1control room and observed the control room operators implement Abnormal ProcedureABN-712, "Rod Control System Malfunction," Revision 10. The inspectors observed theoperators enter the applicable Technical Specification and Technical RequirementManual required actions within the specified completion times, and reduce power, initiallyto 1150 MWe, then less than 75 percent reactor power, and subsequently to less than 50percent reactor power. The inspectors also observed the troubleshooting efforts toidentify the cause of the dropped rod and the briefings between the control room staff andthe maintenance, engineering, and fuel vendor representatives.

-21-The inspectors observed the restoration of Control Rod J13 to its desired position andreviewed the subsequent activities to restore core axial flux difference and quadrantpower flux ratio to acceptable, steady state values.The inspectors interviewed the core performance engineers and assessed the actionstaken by the licensee to control reactor power and core power distributions during theevent and during the restoration to normal full power operations. The inspectors also reviewed the initial troubleshooting results to assess the extent and thoroughness of theefforts to identify the cause of the dropped rod.

b. Findings

No findings of significance were identified.4OA6Meetings, Including ExitExit Meeting SummaryOn October 25, 2007, the inspector presented the occupational radiation safetyinspection results to Mr. M. Kanavos, Plant Manager, and other members of his staff whoacknowledged the findings. The inspector confirmed that proprietary information was notprovided or examined during the inspection.On December 20, 2007, the resident inspection results were presented in a pre-exitbriefing to Mr. M. Blevins, Executive Vice President and Chief Nuclear Officer, and othermembers of Luminant Power management. The inspectors communicated to thelicensee that a final exit would occur upon completion of the inspection. Luminant Poweracknowledged the findings presented.On December 27, 2007, the inspector conducted a telephonic exit meeting withMr. R. Kidwell, Senior Nuclear Analyst, Regulatory Affairs, to verify that no changes to thelicensee's emergency plan or implementing procedures were submitted by the licenseebetween January and December 2007.On January 2, 2008, the inspector conducted a telephonic exit meeting with Mr. T. Hope,Nuclear Licensing Manager, upon completion of the resident inspections. The inspectors asked Luminant Power whether any materials examined during theinspection should be considered proprietary. Proprietary information was reviewed by theinspectors and left with Luminant Power at the end of the inspection.4OA7 Licensee Identified ViolationsThe following violations of very low safety significance (Green) were identified by thelicensee and are violations of NRC requirements which meets the criteria of Section VI ofthe NRC Enforcement Policy for being dispositioned as an NCV:

• Licensee Technical Specifications Section 5.4.1.e requires written procedures beestablished, implemented, and maintained covering all programs in TS 5.5, which

-22-includes TS 5.5.16 Containment Leakage Rate Testing Program. 10 CFR 50Appendix J, paragr aph III.A.1(d) requires that vented syst ems shall be drained ofwater ...to assure exposure of the ... valves to containment air test pressure. Contrary to this requirement, on July 23, 2007, the licensee identified that theprocedure implementing local leak rate testing of containment penetration 2-MIV-0001 did not assure exposure of the valves to air. The piping configurationincluded an N16 delay loop, which acted as a loop seal and prevented the linefrom draining. This piping configuration was identified by the licensee whilereviewing isometric drawings to troubleshoot a leaking containment isolationvalve. The licensee performed a risk evaluation required by TS SR 3.0.3 for asurveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and determined that, although theadministrative leakage limit for this containment isolation valve may have beenexceeded, past failures have resulted in leak rates significantly less than theavailable margin for the overall TS containment leak rate. This condition wasdocumented in SMF-2007-2302-00. This violation is more than minor because itaffected the Barrier Integrity cornerstone objective to assure the containment'sability to protect the public fr om radio nuclide releases, but is very low significance(Green) because it does not increase core damage frequency and, due to thesmall size of the tubing/piping, would not contribute to large early releasefrequency.

• Licensee Technical Specification Section 5.7.1.a. requires that each entryway to high radiation areas not exceeding 1.0 rem per hour be barricaded andconspicuously posted as a high radiation area. Contrary to this requirement, onJuly 26, 2007, after performing a resin transfer, the lower valve gallery wasde-posted from a high radiation area to a radiation area. Upon reviewing thesurvey it was noted that two areas should have remained posted as high radiationareas. This issue was entered into the licensee's corrective action program asSMF-2007-2351-00. This finding is of very low safety significance because it didnot involve a very high radiation area or personnel overexposure.

• Licensee Technical Specification Section 5.7.2. requires that each entryway to high radiation areas exceeding 1.0 rem per hour but less than 500 rads per hourbe barricaded and conspicuously posted as a high radiation area and requires thearea to be conspicuously posted and locked or continuously guarded. TheTechnical Specification also states that if no enclosure can reasonably beconstructed around the area, a clearly visible flashing light shall be activated atthe area as a warning device. Contrary to this requirement, on July 26, 2007, thefuel handling building was not posted as a high radiation area and was neitherguarded or identified with a visible flashing light during transfer of a high integritycontainer to a storage cask. This event was entered into the licensee's correctiveaction program as Smart Form 2007-2322-00. This finding is of very low safetysignificance because it did not involve a very high radiation area or personneloverexposure. This is also considered a Performance Indicator occurrence andwas properly reported with the third quarter performance data. ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Luminant Power personnel

M. Blevins, Executive Vice President and Chief Nuclear Officer

M. Bozeman, Supervisor, Emergency Planning

S. Bradley, Supervisor, Health Physics

A. Caves, ALARA Coordinator, Radiation Protection

R. Fishencord, Sr. Nuclear Specialist, Emergency Planning

R. Flores, Site Vice President

J. Gallman, Senior Nuclear Analyst (Work Week Coordinator)

J. Goodrich, ALARA Technician, Radiation Protection

D. Holland, Senior Nuclear Analyst (Work Week Coordinator)

T. Hope, Nuclear Licensing Manager

M. Kanavos, Plant Manager

S. Karpyak, Risk & Reliability Engineering Supervisor

R. Kidwell, Senior Nuclear Analyst, Regulatory Affairs

B. Kneels, Supervisor, Radiation Protection

B. Knowles, Supervisor, Radiation Protection

G. Krishnan, Procurement Engineering & Program Manager, SHAW

D. Kross, Director, Operations

F. Madden, Director, Regulatory Affairs

S. Maier, Design Engineering Analysis Manager, Technical Support

M. McCutchen, System Engineer

J. Mercer, Maintenance Rule Coordinator

J. Meyer, Technical Support Manager

W. Morrison, Maintenance Smart Team Manager

D. O'Connor, Supervisor, Radiation Protection

B. Patrick, Manager, Radiation Protection

D. Reimer, Manager of Plant Support

J. Rankin, ALARA Technician, Radiation Protection

J. Rincon, ALARA Technician, Radiation Protection

T. Robison, Sr. Nuclear Specialist, Emergency Planning

J. Seawright, Consulting Engineer, Regulatory Affairs

R. Segura, Nuclear Analyst Consultant (Electrical Systems)

S. Smith, Director, System Engineering

D. Sparks, Senior Nuclear Analyst (Work Week Coordinator)

C. Tran, Engineering Programs Manager

D. Wilder, Manager, Security, Emergency Planning, and Environmental

NRC

D. Allen, Senior Resident Inspector

A. Sanchez, Resident Inspector

AttachmentA-2

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

NoneOpened and

Closed

05000445;446/2007005-01NCVFailure to Evaluate Radiological Conditions(Section 2OS1)05000446/2007005-02NCVFailure to Use a Procedure Appropriate to theCircumstances Led to Inadvertent Actuations of theTurbine Driven Auxiliary Feedwater Pump(Section 4OA2.4)

Closed

None

Discussed

None

Section 1R04: Equipment Alignment (71111.04)Unit 1 Train A Residual Heat Removal SystemDrawingsM1-0229, Component Cooling Water System, Revision

CP-22M1-0260, Residual Heat Removal System, Revision

CP-32M1-0263, Safety Injection System, Revision

CP-13Control Room Air Conditioning SystemSTA-759, "Control Room Envelope Habitability Program," Revision 0STA-758, "Ventilation Filter Testing Program," Revision 0OPT-210, "Control Room VAC. System," Revision 9Technical Specifications 3.7.10 Control Room Filtration/Pressurization System (CREFS)Technical Specifications 3.7.11 Control Room Air Conditioning System (CRACS)Current Maintenance Rule Reliability StatusCPNPP System Status Report, 3