IR 05000528/2007011

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May 30, 2007

Randall K. EdingtonSenior Vice President, Nuclear Mail Station 7602 Arizona Public Service Company P.O. Box 52034 Phoenix, AZ 85072-2034

SUBJECT: PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2, 3 - NRCCOMPONENT DESIGN BASES INSPECTION REPORT 05000528/2007011; 05000529/2007011; AND 05000530/2007011

Dear Mr. Edington:

On May 25, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection atyour Palo Verde Nuclear Generating Station, Units 1, 2, and 3. The enclosed report documents the inspection results, which were discussed on March 23, 2007, with Mr. R. Bement, Vice President, Nuclear Operations, and other members of your staff. On May 25, 2007, an exit teleconference was held with Mr. R. Randels, Director, Design Engineering.This 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 cognizant plant personnel. Based on the results of this inspection, the NRC identified four findings that were determined tobe more than minor; three of the findings were determined to be violations of NRC requirements. These findings were evaluated under the risk significance determination process as having very low safety significance (Green). The violations are being treated as noncited violations, consistent with Section VI.A of the Enforcement Policy. If you contest the violations or significance of these noncited violations, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Palo Verde Nuclear Generating Station, Units 1, 2, and 3 facility.On February 20, 2007, the NRC held a public meeting with members of your staff to discuss theprogress of your component design basis review program. During that meeting, your staff discussed the scope of several component reviews that had been performed and documented.

Arizona Public Service Company-2-During this inspection the NRC independently selected several components that had beenreviewed by your staff and noted numerous issues that had not been identified by your staff.

For example, your staff's review of components associated with the station blackout generators did not identify numerous issues, some of which are documented in this report, that should have been identified as part of your independent component design bases review. Based on our sample review, it is evident that your initial effort in conducting component reviews has not been fully effective. We understand that you are implementing actions to improve the component design basis reviews. The NRC will conduct additional inspections at a later date to assess the effectiveness of your actions.In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letterand its enclosure will be available electronically for public inspection in the NRC PublicDocument Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/William B. Jones, ChiefEngineering Branch 1 Division of Reactor SafetyDockets: 50-528, 50-529, 50-530Licenses: NPF-41, NPF-51, NPF-74

Enclosures:

Inspection Report 05000528/2007011; 05000529/2007011; and 05000530/2007011 w/Attachment Supplemental Informationcc w/enclosures:Steve Olea Arizona Corporation Commission 1200 W. Washington Street Phoenix, AZ 85007Douglas K. Porter, Senior CounselSouthern California Edison Company Law Department, Generation Resources P.O. Box 800 Rosemead, CA 91770 Arizona Public Service Company-3-ChairmanMaricopa County Board of Supervisors 301 W. Jefferson, 10th Floor Phoenix, AZ 85003Aubrey V. Godwin, DirectorArizona Radiation Regulatory Agency 4814 South 40 Street Phoenix, AZ 85040Scott Bauer, Acting General ManagerRegulatory Affairs and Performance Improvement Palo Verde Nuclear Generating Station Mail Station 7636 P.O. Box 52034 Phoenix, AZ 85072-2034Jeffrey T. WeikertAssistant General Counsel El Paso Electric Company Mail Location 167 123 W. Mills El Paso, TX 79901John W. SchumannLos Angeles Department of Water & Power Southern California Public Power Authority P.O. Box 51111, Room 1255-C Los Angeles, CA 90051-0100John TaylorPublic Service Company of New Mexico 2401 Aztec NE, MS Z110 Albuquerque, NM 87107-4224Geoffrey M. CookSouthern California Edison Company 5000 Pacific Coast Hwy, Bldg. N50 San Clemente, CA 92672Robert HenrySalt River Project 6504 East Thomas Road Scottsdale, AZ 85251 Arizona Public Service Company-4-Brian AlmonPublic Utility Commission William B. Travis Building P.O. Box 13326 1701 North Congress Avenue Austin, TX 78701-3326Karen O'ReganEnvironmental Program Manager City of Phoenix Office of Environmental Programs 200 West Washington Street Phoenix, AZ 85003 Matthew BenacAssistant Vice President Nuclear & Generation Services El Paso Electric Company 340 East Palm Lane, Suite 310 Phoenix, AZ 85004 Arizona Public Service Company-5-Electronic distribution by RIV:Regional Administrator (BSM1)DRP Director (ATH)DRS Director (DDC)DRS Deputy Director (RJC1)Senior Resident Inspector (GXW2)Branch Chief, DRP/D (RLN1)Senior Project Engineer, DRP/D (GEW)Team Leader, DRP/TSS (FLB2)RITS Coordinator (MSH3)DRS STA (DAP)V. Dricks, PAO (VLD)D. Cullison, OEDO RIV Coordinator (DGC)ROPreports PV Site Secretary (PRC)SUNSI Review Completed: _CJP__ADAMS: O YesG No Initials: __CJP_ O Publicly Available G Non-Publicly Available G SensitiveO Non-SensitiveC:\FileNet\ML071500445.wpdDRS/EB1EB1EB1OBC:EB1DRP/CDD:DRPCPaulkJReynosoJNadelTStetkaWBJonesNO'KeefeAVegel

/RA/ /RA/ /RA/ /RA/ /RA//RA//RA/5/10/075/2/074/24/074/30/075/2/075/10/075/15/07C:EB1WBJones/RA/

5/30/07OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax Enclosure-1-ENCLOSUREU.S. NUCLEAR REGULATORY COMMISSION REGION IV Dockets:50-528, 50-529, 50-530Licenses:NPF-41, NPF-51, NPF-74 Report No.:05000528/2007011; 05000529/2007011; 05000530/2007011 Licensee:Arizona Public Service Company Facility:Palo Verde Nuclear Generating Station, Units 1, 2, and 3 Location:5951 S. Wintersburg Road Tonopah, Arizona Dates:February 19 through May 25, 2007 Team Leader:C. Paulk, Senior Reactor Inspector, Engineering Branch 1 Inspectors:J. Nadel, Reactor Inspector, Engineering Branch 1J. Reynoso, Reactor Inspector, Engineering Branch 1 T. Stetka, Senior Operations Engineer, Operations BranchContractors:C. Baron, Mechanical Engineering Contractor, Beckman & AssociatesS. Kobylarz, Electrical Engineering Contractor, Beckman & AssociatesApproved By:William B. JonesEngineering Branch 1 Division of Reactor Safety Enclosure-2-

SUMMARY OF FINDINGS

IR 05000528/2007011; 05000529/2007011; 05000530/2007011; 2/19/07 - 5/25/07; Palo VerdeNuclear Generating Station, Units 1, 2, and 3: baseline inspection; NRC Inspection

Procedure 71111.21, Component Design Basis Inspection.The report covered a 5-week period of inspection by six region-based inspectors and twocontractors. Three noncited violations and one finding (all Green) were identified. The significance of most findings is indicated by its color (Green, White, Yellow, Red) using Inspection Manual Chapter 0609, "Significance Determination Process." Findings for which the significance determination process does not apply may be green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3, dated July 2000.A.NRC - Identified Findings

Cornerstone: Mitigating Systems

Green.

The team identified a finding involving the implementation of RegulatoryGuide 1.155, Station Blackout, Appendix A, for the demonstration of the station backoutgenerator design and system readiness requirements. Specifically, established preventive maintenance tasks did not demonstrate that the coping requirements for the station blackout generator would be met for the approved increase from the 4-hour to 16-hour coping duration that, at the time this finding was identified, would become effective the following month. The licensee has entered this finding into their corrective action program as Palo Verde Action Request PVAR 2982699.The finding is greater than minor because it would become a more significant safetyconcern if left uncorrected following the implementation of the 16-hour coping duration.

The finding affected the mitigating systems cornerstone attributes to ensure the availability of the station blackout generators to respond to initiating events necessary to prevent undesirable consequences. Using the NRC Inspection Manual Chapter 0609,

Significance Determination Process, Phase 1 Worksheet, the team determined that thisfinding had very low safety significance because there was not a loss of system function and it did not involve an external event. The cause of the finding was related to the crosscutting element of decision making associated with human performance for thefailure to adequately evaluate the design and system readiness requirements for the station blackout generators for the approved license amendment that, at the time the finding was identified, would, increase the coping period to 16-hours.

(Section 1R21b.1.)Green. The team identified a noncited violation of very low safety significance for thefailure to implement the design control requirements of Regulatory Guide 1.155, StationBlackout, Appendix A, Criterion 1, Design Control and Procurement Control, to 10 CFR50.63, Loss of All Alternating Current. Specifically, approved Design ChangeDMWO 2827452 did not account for key station blackout generator performance parameters that included fuel and lubricating oil consumption rates and required station Enclosure-3-blackout battery capacity for an increase in the station blackout coping period from 4to16-hours.The finding is greater than minor because it would become a more significant safetyconcern if left uncorrected in that the critical performance parameters for ensuring the station blackout generators would meet the 16-hour coping requirement were not established. The finding affected the mitigating systems cornerstone attributes to ensure the availability of the station blackout generators to respond to initiating events necessary to prevent undesirable consequences. Using the NRC Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the teamdetermined that this finding had very low safety significance because there was not a loss of system function and it did not involve an external event. The cause of the finding was related to the crosscutting element of decision making associated with humanperformance for the failure to evaluate the key performance parameters for the station blackout generators for the approved license amendment that increased the coping period to 16-hours. (Section 1R21b.2.)Green. The team identified a noncited violation of very low safety significance of 10CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the designcalculation that determined the minimum containment flood level following a loss-of-coolant accident was not based on the most limiting reactor coolant system break location. The calculated containment flood level was used to verify the adequacy of the available net positive suction head for the emergency core cooling pumps that would take suction from the containment sump during the recirculation phase of a postulated loss-of-coolant accident. The licensee has entered this issue into their corrective action program as Palo Verde Action Request PVAR 2981257.This finding is greater than minor because this issue required accident analysiscalculations to be re-performed to assure the accident requirements were met. The finding affected the mitigating systems cornerstone as related to the availability, reliability, and capability of the emergency core cooling system for post-loss-of-cooling accident. In accordance with Inspection Manual Chapter 0609, SignificanceDetermination Process, Appendix A, Significance Determination of Reactor InspectionFindings for At-Power Situations, the team conducted a Phase 1 screening anddetermined the finding was of very low safety significance because it did not represent an actual loss of safety function. This deficiency would not have resulted in the emergency core cooling pumps becoming inoperable under the most limiting postulated accident conditions. This finding has cross-cutting aspects associated with corrective action of the problem identification and resolution area to ensure that issues potentially impacting nuclear safety are promptly identified, fully evaluated and that actions are taken to address safety issues in a timely manner. (Section 1R21b.3.)Green. The team identified a noncited violation of very low safety significance of 10CFR Part 50, Criterion XVI, Corrective Actions, for the failure to identify and correct significant conditions adverse to quality involving Target Rock valve failures. The licensee has entered this issue into their corrective action program as Palo Verde Nuclear Generating Station Action Requests PVAR 2984832 and 2985372.

Enclosure-4-The failure to identify and correct the cause(s) of turbine-driven auxiliary feedwaterpump Target Rock solenoid-operated valves was a performance deficiency. This issue is more than minor because it is associated separately with the mitigating systems cornerstone and on one occasion affected the containment barrier integrity cornerstone.

This finding has cross-cutting aspects associated with corrective action of the problem identification and resolution area to ensure that issues potentially impacting nuclear safety are promptly identified, fully evaluated and that actions are taken to address safety issues in a timely manner. (Section 1R21b.4.)B.Licensee-Identified FindingsViolations of very low safety significance, which were identified by the licensee havebeen reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensee's corrective action program. These violations and corrective actions are listed in Section 4OA7 of this report.

Enclosure-5-

REPORT DETAILS

1.REACTOR SAFETYInspection of component design bases verifies the initial design and subsequentmodifications and provides monitoring of the capability of the selected components and operator actions to perform their design bases functions. As plants age, their design bases may be difficult to determine and an important design feature may be altered or disabled during a modification. The plant risk assessment model assumes the capability of safety systems and components to perform their intended safety function successfully. This inspectable area verifies aspects of the Initiating Events, Mitigating Systems and Barrier Integrity cornerstones for which there are no indicators to measure performance.In addition to performing the baseline inspection, the team reviewed actions taken bythe licensee in response to previously identified significant issues associated with engineering performance.

1R21 Component Design Bases Inspection

The team selected risk-significant components and operator actions for review usinginformation contained in the licensee's probabilistic risk assessment. In general, this included components and operator actions that had a risk achievement worth factor greater than two or Birnbaum value greater than 1E-6.

a. Inspection Scope

To verify that the selected components would function as required, the team revieweddesign basis assumptions, calculations, and procedures. In some instances, the team performed independent calculations to verify the appropriateness of the licensee engineers' conclusions. The team also verified that the condition of the components was consistent with the design bases and that the tested capabilities met the required criteria.The team reviewed maintenance work records, corrective action documents, andindustry operating experience information to verify that licensee personnel considered degraded conditions and their impact on the components. For the review of operator actions, the team observed operators during simulator scenarios associated with the selected components, as well as observing simulated actions in the plant.The team performed a margin assessment and detailed review of the selectedrisk-significant components to verify that the design bases have been correctly implemented and maintained. This design margin assessment considered original design issues, margin reductions due to modification, or margin reductions identified as a result of material condition issues. Equipment reliability issues were also considered in the selection of components for detailed review. These included items such as failed performance test results; significant corrective actions; repeated maintenance; 10 CFR 50.65(a)1 status; operable, but degraded, conditions; NRC resident inspector

-6-input of problem equipment; system health reports; industry operating experience; andlicensee problem equipment lists. Consideration was also given to the uniqueness and complexity of the design, operating experience, and the available defense in depth margins. The inspection procedure requires a review of 15-20 risk-significant and low designmargin components, 3 to 5 relatively high-risk operator actions, and 4 to 6 operating experience issues. The sample selection for this inspection was 20 components, 6 operator actions, and 5 operating experience items. The components selected for review were:

• 13.8kV Bus E-NAN-S03, Breaker S03AB*4.16kV Bus E-PBA-S03, protective relaying

• Containment spray pump

• Emergency Battery B

• Emergency core cooling system relief valves

• Essential cooling water pump

• Essential chilled water pump

• High pressure safety injection bypass Valves SI-698/699

• Low pressure safety inject suction Valve SI-651

• Main steam isolation and atmospheric dump valve accumulators

• Motor-driven auxiliary feedwater pumps

• Refueling water storage tank level and temperature

• Shutdown cooling heat exchanger

• Spray pond level

• Start-up transformer

• Station blackout generator - electrical

• Station blackout generator - mechanical

• Turbine-driven auxiliary feedwater pump casing drains

• Turbine-driven auxiliary feedwater pump steam admission bypass valves

• Turbine-driven auxiliary feedwater pump steam trapsThe selected operator actions were:

• Failure of spray pond pump to start/align auxiliary feedwater pump to refuelingmakeup water tank*Failure of station blackout generator to start

• Align auxiliary feedwater pump n suction manually

• Failure of containment spray pump

• Feed with condensate pumps with a failure to depressurize

-7-The operating experience issues were:*Air-operated valves*Barton transmitters

• Buried cables

• Very low sulfur diesel fuel

• DC-powered motor-operated valves

b. Findings

b.1.Demonstration of Conformance to Design and System Requirements for the Alternateac Power Sources Required for Station Blackout Coping CapabilityIntroduction. The team identified a finding of very low safety significance (Green) for noteffectively demonstrating the station blackout generator conformance with design and system requirements, for the pending increase in the station blackout coping duration, as provided by Regulatory Guide 1.155, Station Blackout, Appendix A, QualityAssurance Guidance and Non-Safety Systems and Equipment.

Description.

Regulatory Guide 1.155, Station Blackout, Section C.3.3.5, states, in part,that the alternate ac power source should have sufficient capacity to operate the systems necessary for coping with a station blackout for the time required to bring and maintain the plant in safe shutdown. Regulatory Guide 1.155, Section 3.5, QualityAssurance and Specification Guidance for Station Blackout Equipment That Is Not Safety-Related, states that the subject guidance is provided in Appendices A, QualityAssurance Guidance for Non-Safety Systems and Equipment; and B, GuidanceRegarding System and Station Equipment Specifications, of the Regulatory Guide. Inresponse to 10 CFR 50.63, Loss of All Alternating Current Power, as stated in ArizonaPublic Service's letter to USNRC, No. 102-05370-CDM/TNW/RAB, dated October 28, 2005, Revised Station Blackout (Station blackout) Evaluation, the licensee adoptedRegulatory Guide 1.155, Sections 3.3.5 and 3.5, and Appendix A, as the manner by which they would meet the requirements of 10 CFR 50.63.

This includes Appendix A, Criterion 5, Testing and Test Control, which contains requirements for a test program toensure that testing is performed to demonstrate conformance with design and system requirements. On October 31, 2006, the NRC approved Amendment 157 to the Palo Verde NuclearGenerating Station operating license. This amendment changed the coping requirement for a station blackout event from 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, and allowed 6 months for the required modifications to be completed before compliance was required. This change affected the design and system readiness requirements of the station blackout system and components.When the Station blackout generators were initially installed, the required copingcapability was 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Pre-operational testing demonstrated the capability of the station blackout generators to supply the design loads for the 4-hour duration.

Subsequent periodic testing has consisted of operating the generators for approximately

-8-1 hour per month and up to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> during each refueling outage (approximately16 hours per year).The team found that the maintenance program established by licensee personnel didnot consistently meet the vendor's requirements/recommendations. For example, many maintenance activities specified in the Palo Verde Nuclear Generating Station PM[preventive maintenance] Program Basis were on a biannual frequency. An exceptionwere the lubricating oil filters which were scheduled for replacement on an 18-month frequency. Licensee personnel established this frequency on the basis of the infrequent operation of the station blackout generators. The vendor, however, identified the filter replacement as a "mandatory" requirement at a frequency of 6 months, "regardless of operating hours." The air filters for the starting air diesel were scheduled for cleaning/replacement on a6-month frequency. However, as observed by the team, the 6-month frequency for the cleaning/replacement of air filters was inadequate. The filters for Train A were replaced on March 7, 2007, after the team noted that they were clogged with dirt and debris. The Train B filters were scheduled for replacement on, or about, April 24, 2007; however, the team also noted that these air filters were partially clogged.Another deficiency the team identified that could affect the capability of the stationblackout generators to run for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> (a design requirement) was the existence of delamination of the internal coating in the fuel oil storage tank. While licensee personnel had noted this condition, there was no established preventive maintenance task. The licensee had elected to perform maintenance on the fuel oil filters on an "as required" basis. Similar concerns were noted for preventive maintenance tasks associated with the testing and inspecting the fuel oil storage tank emergency vent valves, the combination pressure relief/vacuum breaker valves, and fuel oil.Analysis. The team determined that the ineffective demonstration of conformance withdesign and system readiness requirements through effective preventive maintenance was a performance deficiency.While the testing and maintenance program established to demonstrate theconformance with design and system readiness requirements to meet the 4-hour coping requirement was minimal, the ability of the station blackout system to perform its design functions was demonstrated. However, the established testing and maintenance program had not been identified as requiring modification to demonstrate conformance with the design and system readiness requirements to cope for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. As a result, the team found that the existing program would not effectively demonstrate conformance to design and system readiness requirements.The finding is greater than minor because it affects the mitigating systems cornerstoneattributes of design control, equipment performance, procedure quality, and human performance, which affect the cornerstone objective to ensure the availability of systems that respond to initiating events necessary to prevent undesirable consequences. Using the NRC Inspection Manual Chapter 0609, Significance Determination Process, Phase 1Worksheet, the team determined that this finding had very low safety significance

-9-(Green) because there was no loss of system function and it did not involve an externalevent. The licensee has entered this into the corrective action program as Palo Verde Action Request PVAR 2982699.The cause of the finding was related to the crosscutting element of decision makingassociated with human performance for the failure to adequately evaluate the design and system readiness requirements for the station blackout generators for the approved license amendment that increased the coping period to 16-hours.

Enforcement.

No violation of regulatory requirements was identified. This issue isidentified as FIN 05000528, -529, -530/2007011-001, Ineffective Demonstration ofConformance to Design for the Alternate ac Power Sources. b.2.Inadequate Control of Design Information for the Station Blackout SystemIntroduction. A noncited violation of very low safety significance (Green) was identifiedfor the failure to include design-related guidelines used in complying with 10CFR50.63 in design documents.Description. The current station blackout coping requirement of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> was increasedto 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after License Amendment 157 was approved (October 31, 2006). In order to implement the amendment, the Design Change DMWO 2827452 must be implemented. During the review of Design Change DMWO 2827452, the team found that licenseeengineers did not account for key performance parameters. The performance section of the vendor technical manual addresses fuel and lubricating oil consumption as key parameters to be monitored to demonstrate the capability of the station blackout generators to perform their design requirements (i.e., provide an alternate ac power source for the required coping time). The team also found that licensee engineers had not performed battery sizingcalculations to determine the required battery capacity for the station blackout batteries.

To complicate matters, the team found that the battery testing procedures were inadequate to provide an accurate indication of the health of the batteries. In addition, the licensee engineers' review of key structures and components was not adequate to identify or properly classify components (e.g., fuel oil storage tank vent and relief valves)which impact station blackout system operation. Although the station blackout system was identified as requiring an augmented qualityprogram, the team found that licensee personnel did not implement such a program with respect to design control. The licensee has entered these issues into their correctiveaction program as Palo Verde Action Requests PVARs 2980758, 2982699, and 2985197.Analysis. The NRC issued Regulatory Guide 1.155, Station Blackout, as an acceptablemethod for meeting the requirements of 10 CFR 50.63. The Licensee adopted the methodology described in the regulatory guide to comply with 10 CFR 50.63. The team

-10-determined that the failure to control the design information for the station blackoutsystem is a performance deficiency; is more than minor because it is associated with the mitigating system cornerstone attributes of design control, procedure quality, human performance, and equipment performance; and it affected the cornerstone objective to ensure the availability of systems that respond to initiating events necessary to prevent undesirable consequences. Quality assurance, as defined in Regulatory Guide 1.155, Appendix A, requires a qualityprogram for equipment which is used to meet the requirements of 10 CFR 50.63 and not explicitly covered by existing quality assurance requirements. Criterion 1, DesignControl and Procurement Document Control, of Appendix A to Regulatory Guide 1.155states that "[m]easures should be established to ensure that all design related guidelines used in complying with §50.63 are included in design and procurement documents, and that deviation therefrom are controlled."The cause of the finding was related to the crosscutting element of decision makingassociated with human performance for the failure to evaluate the key performance parameters for the station blackout generators for the approved license amendment that increased the coping period to 16-hours. Since no actual loss-of-safety function of the station blackout system has occurred as a result of the inadequate design control, the team determined that this finding was of very low safety significance (Green) in Phase 1 of the significance determination process.Enforcement. Criterion 1, Design Control and Procurement Document Control, ofAppendix A to Regulatory Guide 1.155 states that "[m]easures should be established to ensure that all design related guidelines used in complying with §50.63 are included in design and procurement documents, and that deviation therefrom are controlled."Contrary to the above, as of March 15, 2007, the measures established to ensure thatall the design-related guidelines for the station blackout system were not adequate in that key design parameters were not included in the design documentation used to demonstrate compliance with 10 CFR 50.63 and Regulatory Guide 1.155. Because the finding is of very low safety significance (Green) and has been entered into the licensee's corrective action program, this violation is being treated as a noncited violation, consistent with Section VI.A.1 of the Enforcement Policy: NCV 05000528, -

529, -530/2007011-002, Inadequate Control of Design Information for the SBO System. b.3.Non-conservative Containment Sump Level AnalysisIntroduction. The team identified a noncited violation of 10 CFR 50, Appendix B,Criterion III, Design Control, of very low safety significance for containment flood levelduring certain loss-of-coolant accidents. Specifically, the design calculation that determined the minimum containment flood level following a loss-of-coolant accident was not based on the most limiting reactor coolant system break location. The calculated containment flood level was used to verify the adequacy of the available net positive suction head for the emergency core cooling pumps that would take suction from the containment sump during the recirculation phase of a postulated loss-of-coolant accident.

-11-Description. The team reviewed design Calculation 13-MC-SI-017, Safety InjectionSystem Interface Requirements, Revision 6. In part, this calculation determined theminimum water level that would be available in the containment during the recirculation phase of a postulated loss-of-coolant accident. This calculated water level was used to verify that the emergency core cooling pumps taking suction from the containment sump would have adequate net positive suction head under the most limiting conditions. The team noted that this calculation included a portion of the reactor coolant system inventory in the volume of water that would be available in the containment. The available volume of reactor coolant was based on an assumed break location at, or below, the centerline of the cold leg injection nozzles. The team questioned if this assumed break location was bounding, and if a reactor coolant system break at a higher elevation would result in a lower containment water level during the recirculation phase of a postulated loss-of-coolant accident.In response to this concern, licensee engineers initiated Palo Verde ActionRequest PVAR 2981257 on March 12, 2007. The engineers also issued a prompt operability determination on March 15, 2007. The engineers concluded that there was a reasonable expectation of operability for all the emergency core cooling pumps. They evaluated the potential reduction in emergency core cooling pump net positive suction head margin that would result from this non-conservative analysis input. The design calculations indicated that the limiting emergency core cooling pump net positive suction head margin was 3.8 feet. If the water volume associated with all reactor coolant system spillage was eliminated from Calculation 13-MC-SI-017, the calculated net positive suction head margin would be reduced by less than 1.4 feet. In addition, the engineers evaluated the containment sump screen performance based on a lower water level and concluded that the emergency core cooling system performance would not be adversely affected. The team reviewed the prompt operability determination during the inspection.Analysis. The failure to properly implement design controls was a performancedeficiency. Specifically, design Calculation 13-MC-SI-017 included a non-conservative input value, which affected the available emergency core cooling pump net positive suction head margin under postulated accident conditions. The team determined this finding to be greater than minor because accident analysis calculations were required to be re-performed to assure the accident analysis requirements were met. The finding affected the mitigating systems cornerstone as related to the availability, reliability, and capability of the emergency core cooling system.In accordance with Inspection Manual Chapter 0609, Significance DeterminationProcess, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, the team conducted a Phase 1 screening and determined the findingwas of very low safety significance (Green) because it was a design deficiency confirmed not to result in loss-of-operability in accordance with Part 9900, Technical Guidance, Operability Determination Process for Operability and FunctionalAssessment. Based on the licensee's evaluation, this deficiency would not haveresulted in the emergency core cooling pumps becoming inoperable under the most limiting postulated accident conditions. Licensee personnel entered this into the corrective action program as Palo Verde Action Request PVAR 2981257.

-12-This finding has crosscutting aspects associated with corrective action of the problemidentification and resolution area to ensure that issues potentially impacting nuclear safety are promptly identified, fully evaluated and that actions are taken to address safety issues in a timely manner.

Enforcement:

Criterion III, Design Control, of Appendix B to 10 CFR Part 50 requires, inpart, that measures shall be established for the identification and control of design interfaces and for coordination among participating design organizations. These measures shall include the establishment of procedures among participating design organizations for the review, approval, release, distribution, and revision of documents involving design interfaces. The design control measures shall provide for verifying orchecking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.Contrary to the above, as of March 15, 2007, the design control measures taken werenot adequate to verify that Calculation 13-MC-SI-017, Revision 6, did not include a non-conservative input value which affected the available emergency core cooling pump net positive suction head margin under postulated accident conditions. Because this violation is of very low safety significance and has been entered into the licensee's corrective action program as Palo Verde Action Request PVAR 2981257, this violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000528, -529, -530/2007011-003, Non-conservativeContainment Sump Level

Analysis.

b.4.Inadequate Corrective Actions for Target Rock Solenoid-Operated ValvesIntroduction. The team identified a 10 CFR 50, Appendix B, Criterion XVI, noncitedviolation of very low safety significance (Green) for the failure to promptly identify and correct significant conditions adverse to quality for failures of Target Rock solenoid-operated valves.Description. During the review of the auxiliary feedwater pumps, the team notedrepetitive failures of Target Rock solenoid-valves for the steam admission valves to the turbine-driven pumps. The team also noted Target Rock solenoid-operated valve failures associated with the safety injection tanks and the turbine-driven auxiliary feedwater pump high pressure drain traps.Repetitive failures of the turbine-driven auxiliary feedwater pump steam admissionbypass valves have been a long-standing equipment reliability issue and valve failures have resulted in increased unavailability of the turbine-driven pumps. Corrective actions taken have addressed the symptoms of the failures but have not been effective in addressing the underlying cause. For example, licensee engineers determined, in 2002, that the Target Rock solenoid-valves coils should be replaced every other refueling outage because of accelerated aging from being in a hot environment. Another cause was associated with the tolerances of the solenoid piston ring. A potential contributor was evaluated as the manufacturing tolerance for the piston ring may be too large for the valve applications. Licensee engineers had determined, through maintenance

-13-activities, that by reducing the ring thickness, that the piston ring did not stick and causeblow-by. The blow-by has been postulated as the cause of the heat-related accelerated aging. The valves in question are bolted bonnet, dual pilot assisted, 125Vdc, stainless steelsolenoid-operated valves manufactured by Curtis-Wright Flow Control©. In the applications associated with the turbine-driven auxiliary feedwater pump steam admission bypass line and the safety injection tank system, the valves are normally closed and fail closed on loss-of-power. In the turbine-driven auxiliary feedwater pump high pressure drain system, the valves are normally open and fail closed. The steam admission bypass valves have two safety functions. One is to providecontainment isolation; the other to provide initial steam flow to the auxiliary feedwater turbine-driven pump to warm the steam lines and bring the skid mounted hydraulic control valves and lubrication subsystem to normal operating conditions prior to the larger steam admission valve opening. There are two valves in parallel, each with a steam supply from a different steam generator, in each pump's steam admission line.Over a period dating as far back as 15 years and continuing in current performance,there have been multiple examples of failures and off-normal operation of these steam admission valves for each turbine-driven auxiliary feedwater pump, as well as of similar or identical Target Rock valves in other systems. In most cases, the failures were attributed to one of a small set of known apparent failure causes. A review over the last 2 years of the six steam admission valves onsite, 10 occurrences of valve problems were identified. There were 7 instances of the valves stroking within surveillance acceptance criteria but outside reference values. In a separate occurrence, 1 valve failure resulted in a turbine-driven auxiliary feedwater pump overspeed; and 2 valve failures caused failed surveillance tests, one of which led to a unit shutdown when the valve could not be repaired within its 7-day technical specification allowed action time.

Many of the failures involved sticking piston rings with associated blow-by.Failure of the Target Rock solenoid-operated valves have resulted in significantconditions adverse to quality. However, the licensee has not conducted root cause evaluations for the failures and initiated corrective actions to prevent recurrence. The team noted that a root cause Charter investigation, dated April 26, 2006, was to investigate the cause(s) for unacceptable delays experienced relative to the maintenance and retest of the failed valve, resulting in a unit shutdown. The charter did direct the root cause investigation team "to determine the cause(s) associated with the failure of Valve 2JSGAUV138A, as well as the delays that were experienced in resolving the condition in a controlled and timely manner." However, the root cause investigation team focused on the reasons why maintenance personnel were unable to repair the valve within 7 days, not on the cause of the valve failure or possible corrective actions to prevent recurrence. The root cause investigation did note the historical problems with these valves as the second contributing cause and the failure to replace the coil in accordance with the recommended preventive maintenance frequency as the third contributing cause. Similarly, each time a turbine-driven auxiliary feedwater pump steam admission bypass valve would fail during a surveillance, an apparent cause evaluation would attribute the failure to defects in the Target Rock design.

-14-Analysis. The team found that the failure to identify and correct significant condition(s)adverse to quality, involving the Target Rock solenoid-operated valves was a performance deficiency. The finding was more than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and it affected the cornerstone objective to ensure the availability of systems that respond to initiating events to prevent undesirable consequences. In one instance, the turbine-driven auxiliary feedwater pump Target Rock solenoid-valve failure affected the containment barrier cornerstone. This finding has cross-cutting aspects associated with corrective action of the problem identification and resolution area to ensure that issues potentially impacting nuclear safety are promptly identified, fully evaluated and that actions are taken to address safety issues in a timely manner. Since there was no actual loss of safety function of the pump, the team determined that this finding was of very low safety significance (Green) in Phase 1 of the significance determination process. For the one instance that involved the containment barrier cornerstone, none of the attributes identified in the Manual Chapter 0609, Significance Determination Process, Appendix A, for the containment barrier were affected, and the issue screened as very low safety significance.

Enforcement.

Criterion XVI, Corrective Actions, of Appendix B to 10 CFR Part 50 state,in part, that "[m]easures shall be established to assure that conditions adverse to quality

. . . are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition."Contrary to the above, the the licensee did not promptly identify and correct significantconditions adverse to quality for failures of Target Rock solenoid-operated valves to assure that the cause of the failures were determined and corrective action taken to prevent recurrence. For example, the cause analyses performed for the auxiliary feedwater pump Target Rock solenoid-valve failures, that resulted in a turbine overspeed on one occurrence and a plant shutdown as required by the Technical Specifications, on a separate occurrence, did not promptly identify and provide actions to prevent recurrence. Because the finding is of very low safety significance (Green)and has been entered into the licensee's corrective action program as Palo Verde Action Requests PVARs 2984832 and 2985372, this violation is being treated as a noncited violation, consistent with Section VI.A.1 of the Enforcement Policy: NCV 05000528,

-529, -530/2007011-004, Inadequate Corrective Actions for Target Rock Solenoid-Operated Valves. b.5.Problem Identification and ResolutionAssessment of Corrective Action ProgramInspection ScopeThe team reviewed calculations, drawings, procedures, and other design information forthe components, operator actions, and operating experience items identified above.

Many of those items had also been reviewed by licensee personnel during the

-15-performance of a component design basis review undertaken by the licensee inresponse to previously identified issues associated with engineering performance at the site.The team performed these reviews as part of the inspection procedure, as well as togain an insight into the effectiveness of the licensee's review program and ability to identify conditions adverse to quality. In addition to reviewing the documents, the team performed walkdowns of the selected items and interviewed cognizant plant personnel.AssessmentThe team found that the licensee's component design basis review activities were notcompletely effective in identifying conditions adverse to quality. The team identified several examples of conditions adverse to quality associated with the same components that had also been reviewed by the licensee. (Of these examples, 4 were more than minor and resulted in the noncited violations and findings discussed above.)Based on the team's interviews with licensee personnel and a review of the componentdesign basis review reports, the team was concerned with the thoroughness of the reviews and their understanding of which conditions should be addressed as conditions adverse to quality. Many of the minor violation examples identified by the team had aspects of problem identification deficiencies that were associated with design control and procedural adequacy/implementation. The team was also concerned with an apparent lack of understanding by licensepersonnel of the marginal review program.

Following identification of the issues by the team, licensee personnel promptly initiated appropriate corrective action documents.

Also, as stated above, those examples that were determined to be more than minor were entered into the corrective action programs as Palo Verde Action Requests in accordance with station procedures.4.OTHER ACTIVITIES4OA5Other Activities(Closed) URI 05000528, -529, -530/2005002-04: Potentially Nonconservative SetpointsNRC Inspection Report 05000528, -529, 530/2005002 documented an unresolved itemregarding potentially nonconservative setpoints for safety-related instruments. This item was left unresolved pending review of the licensee's evaluation of these setpoints to determine if there was sufficient margin when all uncertainties were accounted for. The licensee was able to demonstrate that there was sufficient margin in the calculations to demonstrate that the setpoints were conservative. Based on these results, the team identified no performance deficiencies or violations of NRC requirements. This unresolved item is closed.

-16-4OA6Meetings, Including ExitOn March 23, 2007, the team leader presented the inspection results to Mr. R. Bement,Vice President, Nuclear Operations, and other members of the staff who acknowledged the findings. The team leader confirmed that, while proprietary information was provided and examined during this inspection, no proprietary information is included in this report.On May 3, 2007, the team leader presented information related to the classification offindings to Mr. R. Bement, Vice President, Nuclear Operations, and other members of the staff who acknowledged the findings.On May 25, 2007, the team leader presented additional information related to theclassification of findings to an exit teleconference was held with Mr. R. Randels, Director, Design Engineering, and other members of the staff who acknowledged the findings. 4OA7Licensee-Identified ViolationsThe following violations of very low safety significance (Green) were identified bylicensee personnel and are violations of NRC requirements which meet the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as noncited violations.*10 CFR Part 50.55(a)(g)(4), Codes and Standards, states, in part, that ". . .components which are classified as American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 must meet the requirements . . . set forth in Section XI of editions of the ASME Boiler and Pressure Vessel Code." Contrary to this, the licensee identified on January 25, 2007, that required ASME Section XI Inservice Inspections on non-corrosion resistant bolting that was covered by insulation had never been performed. Technical Specifications Surveillance Requirement 3.4.103.1 also requires that these inspections be performed in every 10-year inspection interval per the Code. The affected bolting occurs on valve body-to-bonnet connections and bolted flanges in approximately 50 locations per unit in the safety injection and shutdown cooling systems. This finding is greater than minor because, if left uncorrected, it would lead to a more serious safety concern. Using the Manual Chapter 0609, Phase 1 worksheet, the finding is determined to have very low safety significance (Green)because there was no actual loss of safety function to any component, train, or system. The licensee is currently inspecting the bolted connections and replacing the bolts with corrosion-resistant material. This violation was documented in Palo Verde Action Request PVAR 296298.*10 CFR 50.63(a)(1) requires that a licensed nuclear power plant must be able towithstand and recover from an station blackout event. To meet this requirement, Section 8.3.1.1.10 of the Palo Verde Nuclear Generating Station Updated Final Safety Analysis Report (UFSAR) states that the alternate ac power system is capable of energizing the required loads within one hour of the onset of an

-17-station blackout. The UFSAR also states that a study was performed todemonstrate that Palo Verde Nuclear Generating Station is capable of coping with a station blackout for that initial one-hour period. Contrary to this requirement, the licensee determined, as the result of five tests, that it took from 61 minutes 30 seconds to 67 minutes 30 seconds to energize the required loads.

This issue is documented in the licensee's corrective action program as Palo Verde Action Request PVAR 2970059. This finding is of very low safety significance because testing has demonstrated that, even at the most limiting time of 67 minutes, 30 seconds, Palo Verde Nuclear Generating Station could withstand an station blackout.

A-1

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

G. Andrews, Director, Performance Improvement

R. Bement, Vice President, Nuclear Operations

B. Bolf, Senior Engineer, NSSS System Engineer

M. Brutcher, Section Leader, Design Engineering

R. Buzard, Senior Consultant, Regulatory Affairs

D. Carnes, Director, Nuclear Assurance

C. Churchman, Director, Plant Engineering

G. D'Aunoy, Senior Engineer, PRA Engineering

D. Fan, Department Leader, Special Projects

D. Hautala, Senior Engineer, Regulatory Affairs

J. Hesser, Vice President, Engineering

M. Karbassian, Department Leader, Design Engineering

M. Perito, Plant Manager, Nuclear Operations

R. Randels, Director, Design Engineering

M. Salazar, Section Leader, Maintenance

G. Sowers, Section Leader, PRA Engineering

B. Thiele, Site Manager, Component Design Basis Review

A. Turner, Administrative Assistant, Component Design Basis Review

T. Weber, Section Leader, Regulatory Affairs

J. Wood, Department Leader, Nuclear Training

NRC personnel

T. Brown, Resident Inspector, Diablo Canyon Nuclear Power Plant

J. Melfi, Resident Inspector, Palo Verde Nuclear Generating Station Nuclear

G. Warnick, Senior Resident Inspector, Palo Verde Nuclear Generating Station

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and

Closed

05000528, -529, -530/2007011-01FINIneffective Demonstration of Conformance toDesign for the Alternate ac Power Sources

(Section 1R21b.1.).05000528, -529, -530/2007011-02NCVInadequate Control of Design Information forthe Station Blackout System

(Section 1R21b.2.).

Opened and ClosedAttachmentA-2

05000528, -529, -530/2007011-03NCVNon-conservative Containment Sump LevelAnalysis (Section 1R21b.3.).05000528, -529, -530/2007011-04NCVIneffective Maintenance on Target RockSolenoid-Operated Valves (Section 1R21b.4.).Closed05000528, -529, -530/2005002-04URIPotentially Nonconservative Setpoints(Section 4OA5).

LIST OF DOCUMENTS REVIEWED

Calculations:NUMBERTITLEREVISION02-EC-MA-0221AC Distribution12

2-EC-PB-0200AC Overcurrent Protection: Class 1E8

2-EC-PK-0207DC Battery Sizing and Minimum Voltage Calculation 6

13-EC-NA-0221Non-Class 1E 13.8 kV Switchgear Protection113-EC-PB-0204AC Equipment Protection (4.16KV and 480V): Class 1E413-EC-PH-0100A.C. Power feeder, voltage drop and cable sizeverification1213-ES-A15Station Blackout Coping Study1 13-JC-AF-0205 Turbine Driven AF Pump Control Settings 4

13-JC-CH-0206Refueling Water Tank Level Instruments (CHA-L-200 andCHBL-201) Setpoint and Uncertainty Calculation713-JC-CT-0200Setpoints and Total Loop Uncertainty for High/LowCondensate Tank Levels (Loops CTALLOOPOO35 and

CTBLLOOP0036)9

Calculations:NUMBERTITLEREVISIONAttachmentA-313-JC-DF-0202Diesel Fuel Oil Storage Tank Level Instrument UncertaintyCalculation613-JC-DF-0203Diesel Fuel Oil Storage Tank Level Instrument Setpointand Uncertainty Calculation413-JC-DF-0204Diesel Fuel Oil Storage Tank Level Indication ConversionCalculation213-JC-DG-0201Diesel Fuel Oil Day Tank Level Instruments(DGN-L-5/6,-1/2, -343/344) Uncertainty Calculation313-JC-DG-0203Emergency Diesel Generator (DG) and Diesel Fuel Oil(DF) Systems Instrumentation Uncertainty Calculation713-JC-RC-0202RCS Hot & Cold Leg Temperature InstrumentRCx-T-0112x & Rcx-T-0122x) Uncertainty Calculation813-JC-RC-0204Pressurizer Level Instrument (RCA-L-110X &RCB-L-110Y) Setpoint and Uncertainty Calculation613-JC-SI-0218 Containment Spray Header Water Level Loop Setpointand Uncertainty 313-JC-ZZ-0201MOV thrust, Torque and Actuator Sizing Calculation1013-MC-AF-0209 Turbine Driven AF Pump Warming Line SizingCalculation 413-MC-AF-0210 Turbine Driven AF Pump Response Time 113-MC-AF-0309AF Hydraulic Calculation for Q-Trains7

13-MC-CT-0205Condensate Storage Tank0, 1, 2, 3,413-MC-CT-0307Condensate Storage Tank Minimum Level Setpoint2, 3, 4

13-MC-DF-0302Diesel Fuel Oil Excess Flow Check Valve and InstrumentTubing Leakage Rate1

Calculations:NUMBERTITLEREVISIONAttachmentA-413-MC-DF-0305Calculation of Diesel Generator Fuel Oil Piping Flows,Pressures, and Temperatures013-MC-DF-0306As Built Calculation for Sizing the Diesel Fuel Storage andDay Tanks713-MC-EC-0200EC System Hydraulic Calculation513-MC-EC-0252EC System Water Requirements and Chiller Sizing8

13-MC-HA-0802Auxiliary Building Turbine Driven AFW Pump Room -Temperature Transient Blackout213-MC-SG-0211 AOV Thrust and Actuator Sizing Calculation-CCI DragValves 213-MC-SG-0314Nitrogen Tank Pressure Requirements for ADVs613-MC-SG-0405ADV Nitrogen Tank Temperature Adjusted Pressures2

13-MC-SI-0017Safety Injection System Interface Requirements6

13-MC-SI-0018Containment Spray System Interface RequirementsCalculation713-MC-SI-0021HPSI Orifice Two-Stage Design113-MC-SI-0220Containment Spray System Hydraulic analysis and pumpfull flow and miniflow surveillance test requirements. 313-MC-SI-0222 HPSI Hot Leg Injection MOVs- Maximum Diff. Pressure 213-MC-SP-0306MINET Hydraulic Analysis of SP System4

13-MC-SP-0307SP/EW System Thermal Performance Design BasisAnalysis713-MC-ZZ-0216 Air Operated Valve Bench Set Calculation 613-MC-ZZ-0217Gate Valve Open Thrust Required During PotentialPressure Locking Conditions4

Calculations:NUMBERTITLEREVISIONAttachmentA-513-MC-ZZ-2219 Piston AOV Thrust and Actuator Sizing Calculations 413-NC-SI-0202Containment Spray Initiation Times 3

13-NC-SP-0006Volume of Water in Essential Spray Pond2

13-NC-SP-0202Loops L-27 & L-28 Essential Spray Pond LevelUncertainty and Setpoint Calculation413-NC-SP-0206Ultimate Heat Sink Design Reverification413-NC-ZC-0232Effects of Containment Spray Setpoint on ContainmentPeak Pressure Analysis 973DP-9ZZ14Surveillance Testing 14A0-MA-GT-944Gas Turbine Fuel Oil Temperature0

A0-MC-FS-0201GTG fuel usage during 16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> blackout: Process Levels-Setpoints-Supply line losses and vent sizing 1A0-MC-FS-0500 Fuel Oil System-Station Blackout Piping 1AO-EC-NA-0422Phase and Ground Overcurrent Relay/Breaker Selectionsand Settings for Station Blackout System Protection Devices1S-06-0331Maintenance of Atmospheric Dump Valve Rupture Disc0Corrective Action Documents:CRAI 0077197CRAI 0117961

CRAI 2570696

CRAI 2862270

CRAI 2862295

CRAI 2862298

CRAI 2862303

CRAI 2880505

CRAI 2880515

CRAI 2884612CRAI 2886161CRAI 2913077

CRAI 2913844

CRAI 2921186

CRAI 2921217

CRAI 2930022

CRAI 2932459

CRAI 2935818

CRAI 2938079

CRAI 2938116CRAI 2945501CRAI 2945507

CRAI 2969114

CRAI 2969115

CRAI 2970984

CRAI 2970991

CRDR 0077194

CRDR 0101883

CRDR 0116569

CRDR 0117203

AttachmentA-6CRDR 0117666CRDR 0220394

CRDR 0950242

CRDR 0950297

CRDR 1000043

CRDR 2250970

CRDR 2325514

CRDR 2407009

CRDR 2412347

CRDR 2425950

CRDR 2428211

CRDR 2505473

CRDR 2533249

CRDR 2534537

CRDR 2548716

CRDR 2556817

CRDR 2559098

CRDR 2564721

CRDR 2584124

CRDR 2612302

CRDR 2669828

CRDR 2720228

CRDR 2759239

CRDR 2759581

CRDR 2761657

CRDR 2784074

CRDR 2784303

CRDR 2790508

CRDR 2791162

CRDR 2792214

CRDR 2812118

CRDR 2818836

CRDR 2822409

CRDR 2825638

CRDR 2825644

CRDR 2825647

CRDR 2827845

CRDR 2839237

CRDR 2840390

CRDR 2841701

CRDR 2867216

CRDR 2869959

CRDR 2870352

CRDR 2872073

CRDR 2872154

CRDR 2875982

CRDR 2881083

CRDR 2883283CRDR 2883327CRDR 2886558

CRDR 2897810

CRDR 2898586

CRDR 2898613

CRDR 2902154

CRDR 2905638

CRDR 2906728

CRDR 2908954

CRDR 2909725

CRDR 2913728

CRDR 2916323

CRDR 2921844

CRDR 2928626

CRDR 2932120

CRDR 2932177

CRDR 2934345

CRDR 2935232

CRDR 2936216

CRDR 2936461

CRDR 2938366

CRDR 2945347

CRDR 2952147

CRDR 2958450

CRDR 2970059

CRDR 2971188

CRDR 2971202

CRDR 2971616

CRDR 2976608

CRDR 2981485

CRDR 6694222

EER 89-DG-116

EER 90-AF-011

EER 93W-FS-2545

PVAR 0117960

PVAR 00804911

PVAR 2952316

PVAR 2962981

PVAR 2968754

PVAR 2975783

PVAR 2975917

PVAR 2976608

PVAR 2976688

PVAR 2977447

PVAR 2977456

PVAR 2978124

PVAR 2978127

PVAR 2978156PVAR 2978191PVAR 2978448

PVAR 2979101

PVAR 2979140

PVAR 2979453

PVAR 2979643

PVAR 2979915

PVAR 2980300

PVAR 2980651

PVAR 2980677

PVAR 2980726

PVAR 2980758

PVAR 2980772

PVAR 2980805

PVAR 2981017

PVAR 2981226

PVAR 2981234

PVAR 2981257

PVAR 2981891

PVAR 2982023

PVAR 2982134

PVAR 2982193

PVAR 2982244

PVAR 2982699

PVAR 2982716

PVAR 2982775

PVAR 2982823

PVAR 2982828

PVAR 2982829

PVAR 2983146

PVAR 2983315

PVAR 2983319

PVAR 2983778

PVAR 2984000

PVAR 2984054

PVAR 2984055

PVAR 2984140

PVAR 2984313

PVAR 2984315

PVAR 2984322

PVAR 2984328

PVAR 2984367

PVAR 2984400

PVAR 2984434

PVAR 2984743

PVAR 2984745

PVAR 2984832

PVAR 2984872

AttachmentA-7PVAR 2984912PVAR 2985156

PVAR 2985161PVAR 2985167PVAR 2985197

PVAR 2985242PVAR 2985372Design Basis Manuals:TITLEREVISIONAuxiliary Feedwater16

Diesel Generator, Class 1E Standby Generation, Fuel Oil Storage and TransferSystem16Diesel Generator and Class 1E Generation System1

Essential Cooling Water System18

Essential Spray Pond System15

Station Blackout Gas Turbine Generation System7

Class 1E 4.16 KV Power System7

Class 1E 125 VDC Power System11Diesel Fuel Oil Samples:NUMBERTIME (LOCAL/24 HR)DATEC05-10501115January 5, 2005

C05-10511315January 7, 2005

C05-10620840January 12, 2005

C05-10760820January 25, 2005

C05-11071310February 8, 2005

C05-11120850February 10, 2005

C05-11131035February 10, 2005

C05-11140835February 11, 2005

Diesel Fuel Oil Samples:NUMBERTIME (LOCAL/24 HR)DATEAttachmentA-8C05-11430845February 24, 2005C05-11710810March 10, 2005

C05-11841105January 25, 2006

C05-12110755March 19, 2005

C05-12160840March 24, 2005

C05-121911:05March 29, 2005

C05-12200750February 8, 2006

C05-12708:35April 7, 2005

C05-127213:20April 11, 2005

C05-127315:40April 11, 2005

C05-127421:10April 11, 2005

C05-127522:50April 11, 2005

C05-12760025April 12, 2005

C05-12770835April 12, 2005

C05-12781150April 12, 2005

C05-127914:25April 12, 2005

C05-128020:15April 12, 2005

C05-128121:25April 12, 2005

C05-12820025April 13, 2005

C05-12931520April 18, 2005

Diesel Fuel Oil Samples:NUMBERTIME (LOCAL/24 HR)DATEAttachmentA-9C05-12941345April 18, 2005C05-13000840April 21, 2005

C05-13011225April 21, 2005

C05-13041130April 24, 2005

C05-13051335April 24, 2005

C05-13062020April 24, 2005

C05-13072235April 24, 2005

C05-13080105April 25, 2005

C05-13090800April 25, 2005

C05-13101000April 25, 2005

C05-13111240April 25, 2005

C05-13122020April 25, 2005

C05-13132350April 25, 2005

C05-13140155April 26, 2005

C05-13292235April 26, 2005

C05-13341030May 3, 2005

C05-13451235May 8, 2005

C05-13461410May 9, 2005

C05-14230830May 26, 2005

C05-14300855May 25, 2005

Diesel Fuel Oil Samples:NUMBERTIME (LOCAL/24 HR)DATEAttachmentA-10C05-14480825June 2, 2005C05-14680850June 16, 2005

C05-15400845June 30, 2005

C05-15410830June 30, 2005

C05-15800830July 20, 2005

C05-15840820July 26, 2005

C05-16000845July 28, 2005

C05-16161040August 10, 2005

C05-16191245August 11, 2005

C05-16201325August 11, 2005

C05-16390840August 25, 2005

C05-16421000August 26, 2005

C05-16690830September 8, 2005

C05-16701045September 9, 2005

C05-16950835September 23, 2005

C05-17491125October 6, 2005

C05-17611320October 13, 2005

C05-17950845October 28, 2005

C05-18060845November 3, 2005

C05-18081120November 6, 2005

Diesel Fuel Oil Samples:NUMBERTIME (LOCAL/24 HR)DATEAttachmentA-11C05-18090835November 7, 2005C05-18101130November 7, 2005

C05-18240915November 17, 2005

C05-18350945November 18, 2005

C05-18360815November 18, 2005

C05-18380800November 21, 2005

C05-18390920November 21, 2005

C05-18401540November 21, 2005

C05-18810855December 13, 2005

C05-18850925December 15, 2005

C05-19160850December 28, 2005

C05-19250800December 29, 2005

C06-11600830January 11, 2006

C06-11850840January 26, 2006

C06-11881125January 27, 2006

C06-12221325February 9, 2006

C06-12770820February 23, 2006

C06-13260830March 9, 2006

C06-13510835March 22, 2006

C06-14230740April 12, 2006

Diesel Fuel Oil Samples:NUMBERTIME (LOCAL/24 HR)DATEAttachmentA-12C06-14240945April 12, 2006C06-14251150April 12, 2006

C06-14261325April 12, 2006

C06-14271535April 12, 2006

C06-14281957April 12, 2006

C06-14292150April 12, 2006

C06-14300805April 13, 2006

C06-14310930April 13, 2006

C06-14321130April 13, 2006

C06-14351330April 14, 2006

C06-14362035April 14, 2006

C06-14601035April 20, 2006

C06-14740840April 23, 2006

C06-14751020April 23, 2006

C06-14761315April 23, 2006

C06-14770810April 24, 2006

C06-14780945April 24, 2006

C06-14791125April 24, 2006

C06-14801345April 24, 2006

C06-14811545April 24, 2006

Diesel Fuel Oil Samples:NUMBERTIME (LOCAL/24 HR)DATEAttachmentA-13C06-14820910April 25, 2006C06-14831055April 25, 2006

C06-14841255April 25, 2006

C06-14991515April 26, 2006

C06-15020930May 1, 2006

C06-15220855May 5, 2006

C06-15230825May 4, 2006

C06-15281020May 5, 2006

C06-15311020May 6, 2006

C06-15371015May 10, 2006

C06-15650825May 18, 2006

C06-15661035May 18, 2006

C06-15670845May 18, 2006

C06-16140900June 6, 2006

C06-16231325June 9, 2006

C06-16240935June 13, 2006

C06-16281305June 14, 2006

C06-16331005June 17, 2006

C06-16341255June 17, 2006

C06-16480805June 25, 2006

Diesel Fuel Oil Samples:NUMBERTIME (LOCAL/24 HR)DATEAttachmentA-14C06-16710740June 28, 2006C06-16720902June 28, 2006

C06-16990830July 13, 2006

C06-17030840July 18, 2006

C06-17101220July 21, 2006

C06-17150835July 25, 2006

C06-17161010July 25, 2006

C06-17790855August 3, 2006

C06-17860740August 9, 2006

C06-17890930August 10, 2006

C06-18090915August 24, 2006

C06-18101105August 24, 2006

C06-18170835August 29, 2006

C06-18310855September 5, 2006

C06-18690945September 19, 2006

C06-18740750September 20, 2006

C06-19060810September 27, 2006

C06-19380810October 11, 2006

C06-19420925October 13, 2006

C06-19461240October 15, 2006

Diesel Fuel Oil Samples:NUMBERTIME (LOCAL/24 HR)DATEAttachmentA-15C06-19550900October 24, 2006C06-19630920October 29, 2006

C06-19830800November 2, 2006

C06-19840830November 2, 2006

C06-19940840November 15, 2006

C06-20110850November 22, 2006

C06-20120915November 28, 2006

C06-20620850December 15, 2006

C06-20750755December 15, 2006

C06-20850825December 19, 2006

C06-21041015December 27, 2006

C06-21050845December 27, 2006

C06-21060835December 28, 2006

C07-11570855January 11, 2007

C07-11750900January 25, 2007

C07-12010910February 8, 2007

C07-12300845February 21, 2007

C07-12320905February 22, 2007

AttachmentA-16Drawings:NUMBERTITLEREVISION01-E-PBA-001Single Line Diagram, 4.16 KV Class 1E Power SystemSwitchgear 1E-PBA-S03601-E-PKA-005Single Line Diagram, 125V DC Class 1E Power SystemDC Control Center 1E-PKB-M421001-J-DFL-001Control and Logic Diagram DGFO Pumps and SystemAlarms 201-M-AFP-001P. & I. Diagram - Auxiliary Feedwater System3401-M-CTP-001P. & I. Diagram - Condensate Storage and TransferSystem1901-M-DGP-001P. & I. Diagram - Diesel Generator System4801-M-ECP-001P&I Diagram Essential Chilled Water System31

01-M-EWP-001P. & I. Diagram - Essential Cooling Water System30

01-M-SPP-001P. & I. Diagram - Essential Spray Pond System40

01-P-ZYA-019Reactor Make-Up Water Area Piping Plan & Section0

2-E-PKA-005Single Line Diagram, 125V DC Class 1E Power SystemDC Control Center 2E-PKB-M42702-M-ECP-001P&I Diagram Essential Chilled Water System2902-M-SIP-001P&I Diagram Safety Injection & Shutdown CoolingSystem3502-M-SGP-001P&I Diagram Main Steam System5803-E-PKA-005Single Line Diagram, 125V DC Class 1E Power System,DC Control Center 3E-PKB-M42603-M-AFP-001P&I Diagram Auxiliary Feedwater System2203-M-ECP-001P&I Diagram Essential Chilled Water System22

Drawings:NUMBERTITLEREVISIONAttachmentA-1710407-13-MM-105(2 of 2)Diesel Fuel Oil Storage Tanks810407-13-MM-105(1 of 2)Diesel Fuel Oil Storage Tanks812-P-ZYA-07Diesel Oil Storage Tank Area Piping Plan and Sections1113-E-MAA-001Main Single Line Diagram21

13-E003-00015OA/FOA/FOA Transformer Control Schematic WiringDiagram1413-J-03K-084Reactor Makeup Water Tank413-M-DFP-001P. & I. Diagram - Diesel Fuel Oil and Transfer System17

469621A-C149760,Sheet 26T-G #1 Electrical SchematicB469621A-C149760,Sheet 6T-G #1 Electrical SchematicD469621A-C149760,Sheet 9T-G #1 Electrical SchematicAA0-104-W311-96Draw-off sump- Double Nozzle Sump 1A0-C-ZVC-186Station Blackout Gas Turbine Generator Site Plan 3

A0-E-NAA-006Single Line Diagram, Station Blackout Gas TurbineGenerator Switchgear

AE-NAN-S072A0-E-NAB-024Elementary Diagram, 13.8KV Non-Class 1E PowerSystem, Station Blackout GTG Bus AE-NAN-S07

Feeder Breakers for Units 1, 2, & 32A0-E-NAB-025Elementary Diagram, Stand-By Generation SystemGas Turbine Generators

AE-NEN-G01A & B 13.8KV

Breaker1

Drawings:NUMBERTITLEREVISIONAttachmentA-18A0-EN609-A08413.8KV System One Line Diagram5A0-EN609-A114-5GTG Lube Oil Tank Drawing5

A0-M-GTP-001Station Blackout P&I Diagram 2

A0-P-ZYA-093Station Blackout Gas Turbine Generator UndergroundPiping Plan 2DS-C-61167Nozzle Type Relief ValveCDS-C-61167Nozzle Type Relief Valve0

DS-C-61169-1Nozzle Type Relief Valve0

DS-C-61169-2Nozzle Type Relief ValveC

DS-C-61170-1Nozzle Type Relief Valve0

DS-C-61173-1Nozzle Type Relief Valve0

DS-C-61181Nozzle Type Relief ValveH

SDOC M105-00025Diesel Fuel Oil Storage Tanks11

SDOC M105-00024Diesel Fuel Oil Storage Tanks14

SPEC-13-MM-0105Diesel Fuel Oil Storage Tanks5Miscellaneous:NUMBERTITLEREVISION /DATEMemo 280-1762-MAR, CRDR Action 950169.04Complete Actions 2, 3, and 4June 7, 1995E-mail Smith to Buzard, EDG L.O. ConsumptionRateMarch 6, 2007

Miscellaneous:NUMBERTITLEREVISION /DATEAttachmentA-19E-mail Borrero to Murphy, RWT TemperatureFebruary 27,2007E-mail Smith to Buzard, Max Lube OilMarch 6, 2007

E-mail Smith to Buzard, Max Turbo Oil PressureMarch 6, 2007

E-mail Hodgkins to Bressett, Generator BearingLevel GaugesMarch 5, 2007Component Design Basis Review, Mini Report forPK Components, 125 VDC Class 1E Battery

1EPKBF12 & 14, Train B0PRA Risk Assessment of missed inspection ofASME Class 1, 2, 3 Bolted Connections without Corrosion Resistant Bolting 1Unit 1 Inservice Inspection Report, 12th RefuelingOutage0Unit 2 Inservice Inspection Report, 13th RefuelingOutage0Unit 3 Inservice Inspection Report, 12th RefuelingOutage0Operator BurdensFebruary 12,2007Unit 1 2 3 Night OrderMarch 22,2007102-02300(File: 92-001-419.8)Response to NRC comments on Periodic Testing ofAlternate AC (AAC) Sources. APS letter to NRC. October 2,1992102-02440(File: 93-056-026)Response to NRC comments on Periodic Testing ofAlternate AC (AAC) Sources. APS letter to NRC. March 8, 199313-J-083-034Data Sheet Process Solenoid Valves 3

Miscellaneous:NUMBERTITLEREVISION /DATEAttachmentA-2013-J-083-062Data Sheet Process Solenoid Valves 113-J-083-078Data Sheet Process Solenoid Valves 1

13-JM-603Material Specification-Nuclear Service SolenoidValves 1813-JN-0699Specification Class 1 Solenoid Valves213-NS-B062At-power PRA Study for Human Reliability Analysis6

13-NS-C083Appendix E, Human Actions List by CDF RAWImportance113-SM-AF-042Design Change

SGA-UV-134A &138A installation2161-04146(File: 91-0560-026)Revised Response to the Station Blackout Rule (10CFR 50.63) APS letter to NRCAugust 31,1991161-04684(File: 92-056-026)Response to the NRC Station Blackout SafetyEvaluation.

APS letter to NRC. March 20,1992294-01941-DWVCompany Correspondence - Review of OperabilityDeterminations carried over into 2C14November 17,2006324-00129-CDC(File: 91-014-00)Operations Participation in Station Blackout (SBO) -APS LetterJune 12, 1991AF - AuxiliaryFeedwaterSystem Health ReportJanuary 1 -June 30, 2006APS 161-03025(File: 90-056-026) Supplemental Information on Station Blackout, APSletter to

NRC.March 26,1990APS 161-01842 (File:

89-056-026)Response to the Station Blackout Rule. APS letterto NRCApril 14, 1989CH - Chemical andVolume ControlSystem Health ReportJanuary 1 -June 30, 2006

Miscellaneous:NUMBERTITLEREVISION /DATEAttachmentA-21DG - DieselGeneratorsSystem Health ReportJanuary 1 -June 30, 2006E003-29-1Westinghouse Electric CorporationReport of Transformer Tests, S/N 7002700October 18,1979EW - EssentialCooling WaterSystem Health ReportJanuary 1 -June 30, 2006File: 93-056-026(TAC M68579)Station Blackout Supplemental Safety EvaluationPVNGSApril 14, 1993File: 92-014-000(TAC M68579)Supplemental Station Blackout Safety EvaluationPVNGSJuly 28, 1992File: 93-014-000(TAC M68579)Station Blackout Supplemental Safety EvaluationPVNGSJanuary 4,1993File: 92-056-026Station Blackout Safety Evaluation PVNGSFebruary 11,1992Generic Letter 88-14Instrument Air Supply System Problems AffectingSafety-Related Equipment for PVNGS. Letter to

NRCJanuary 3,1991GT - Gas TurbineGeneratorsSystem Health ReportJanuary 1 -June 30, 2006GTG1.02GTG Test RecordOctober 24,1993IEEE Std 1106IEEE Recommended Practice for Installation,Maintenance, Testing, and Replacement of Vented Nickel-Cadmium Batteries for Stationary Applications1995 & 2005IN 06-14,Supplement 1Potentially Defective External Lead-WireConnections in Barton Pressure TransmittersSeptember 25,2006

Miscellaneous:NUMBERTITLEREVISION /DATEAttachmentA-22IN 06-14Potentially Defective External Lead-WireConnections in Barton Pressure TransmittersJuly 10, 2006LT - LargeTransformersSystem Health ReportJanuary 1 -June 30, 2006MME 02252Material Engineering Evaluation - Target RockCorporation SOV High Temperature Coil P/N

303703-1, Part Substitution Evaluation2NA - Non-Class 1E13.8KV PowerSystem Health ReportJanuary 1 -June 30, 2006NRC RegulatoryIssue Summary

2000-03Performance of Safety Related Power OperatedValves Under Design Basis ConditionsMarch 15,2000NUMARC 87-00Station Blackout Coping Duration-Guidelines andTechnical Bases 1PB - Class 1E4.16KV PowerSystem Health ReportJanuary 1 -June 30, 2006PRIME Engineering Report No. R3-764-79Connector P/N 0764-1221B Design ChangeAcceptabilityJune 23, 2006Regulatory Guide1.155Station Blackout August 1988.RSS-02-16812ENANS03AB, Non-Class 1E, 13.8kV, 51/51NStation Blackout Feed to 2ENANS03A and

2ENBNX03 ESF XFMR3RSS-03-00443ENANS03A, Non-Class 1E, 13.8kV, Cubicle A,Feed to ESF Service Transformer, 3-E-NBN-X03,

10/12.5MVA, 13.8-4.16kV1

Miscellaneous:NUMBERTITLEREVISION /DATEAttachmentA-23RSS-AO-1678AENANS07D, Non-Class 1E, 13.8kV StationBlackout Feed to 1ENANS03 and 1ENBNX03 ESF

Transformer4RSS-AO-1679AENANS07E, Non-Class 1E, 13.8kV Station Blackout Feed to 2ENANS03 and 2ENBNX03 ESF

Transformer4RSS-AO-1680AENANS07F, Non-Class 1E, 13.8kV Station Blackout Feed to 3ENANS03 and 3ENBNX03 ESF

Transformer4S-05-013910

CFR 50.59 Screening - The proposed activitywill upgrade twenty four non-qualified plant instruments and their associated lines and supports to the quality class Q.0S-05-025710

CFR 50.59 Screening - Revision to 73ST-9SG05ADV Nitrogen Accumulator Drop Test 0S-05-0442Safety Evaluation for Calculation 13-MC-CT-307Revise

Minimum CST Level Setpoint to Prevent Air Entrainment0S-06-033110

CFR 50.59 Screening - Add new vendordocuments to support maintenance of ADV Rupture Disc0S-06-039210CFR 50.59 screening-GTG procedure changesSeptember 14,2006S-07-001710

CFR 50.59 Screening - Implement modificationto install (1) a supplemental nitrogen supply to the nitrogen accumulator for each Atmospheric Dump Valve (ADV) and (2) a two-way radio system to facilitate communication between operators in the

GTG control room and operators in the Unit 1, 2

and 3 main control rooms.0SG-1039Design Change Request, 16 Hour Station BlackoutCoping Modification November 16,2006

Miscellaneous:NUMBERTITLEREVISION /DATEAttachmentA-24SP - Spray PondSystem Health ReportJanuary 1 -June 30, 2006Startup Field Report Job No. 10407Battery Charger / Power ConversionAugust 11,1983VTD-A160-0103Allen-Bradley Power Supply Modules InstallationData (Cat. No. 1771-P3, -P4 and -P5) [Pub. #

1771-2.111]2WSL

MI 2304213Instructions for Testing and Calibrating the StartupTransformer Instruments and Performing the Large Transformer Internal CT TestingMarch 25,2004WSL 244336Recharge Task Class 1E Battery "PK" Following aSurveillance Battery Test DischargeOctober 28,2005Modifications:NUMBERTITLEDATEDCR

SG-103916 Hour Station Blackout Coping ModificationsNovember 16,2006DFWO 2882666Relocate Valve SI UV0651September 6,2006DMWO 806536Replace the 4 Class 1E Batteries (AT&T RoundCells) and Battery Racks with Rectangular Batteries and New Racks in All 3 Units Unit 1: January 2,2002Unit 2: May 21,1999Unit 3: June 1,2000DMWO 2827452Changes needed to extend PVNGS' ability to copewith a Station Blackout for a 16-hour loss of all alternating current power, as committed to the NRC

via APS letter No. 102-05370, dated October 28,

20050, Pen & Ink 1

Modifications:NUMBERTITLEDATEAttachmentA-25EDC 2006-00315Engineering Document Change - Doc No. 01-EC-MA-0221AWO 00693599AF Turbine Steam Supply ModificationFebruary 19, 19951-SM-EC-002Setpoint ChangeMarch 14, 1988Operability Determinations:NUMBERTITLEREVISIONPOD for PVAR 29629810

POD for PVAR 29812570285POD for WO#

2762951, Water Intrusion into Diesel StorageTank0305U2 Essential Pipe Chase Seepage0CRDR 2719463Evaluation of Operability Impact to A Train AuxiliaryFeedwater Pump when M23 and M24 Steam Traps are Out of Service - Supplement 10Procedures:NUMBERTITLEREVISION /DATE111-00606-MLH(File: 94-0140-

000)Closure of Station Blackout Test Procedure. APS letterFebruary 23,1994240ST-9EC03Essential Chilled Water & Ventilation SystemsInoperable Action Surveillance1232ST-9PK0460-Month Surveillance Test of Station Batteries27

Procedures:NUMBERTITLEREVISION /DATEAttachmentA-2633TI-9EC01Essential Chilled Water System Flow Balance136MT-9SG01ADV Bonnet Cavity Pressure and InstrumentInstallation639DP-9ZZ02Air Operated Valve Program 1040AL-9MA01Transformer Trouble Alarm Responses23

DP-9AP08Technical Guideline for LOCA

40DP-9OP06Operations Department Repetitive Task Program91

40DP-9OP08Diesel Generator Test Record43

40DP-9OP26Operability Determination and Functional Assessment18

40DP-9OPA4Area 4 Operator Logs, Mode 1-475

40EP-9EO03Loss of Coolant Accident22

40EP-9EO04Steam Generator Tube Rupture20

40EP-9EO05Excess Steam Demand19

40EP-9EO06Loss of all Feedwater13

40EP-9EO07Blackout8

40EP-9EO08Functional Recovery27

40EP-9EO09Standard Appendices53

40EP-9EO10Lower Mode Functional Recovery17

40EP-9EO11Loss of Coolant Accident22

40OP-9AF01Essential Auxiliary Feedwater System36

Procedures:NUMBERTITLEREVISION /DATEAttachmentA-2740OP-9EC01Essential Chilled Water Train "A" (EC)840OP-9NA0313.8 kV Electrical System (NA)23

40OP-9SG01Main Steam50

40ST-9ZZM1Operations Mode 1 Surveillance Logs44

41AL-1RK2AWindow No. 2A07A - ESS CHLD WTR SYS TRBL48

41AL-1RKGACST Empty44

41ST-1EC01Essential Chilled Water Valve Verification17

2AL-2RK1BPanel B01B Alarm Responses24

2ST-2EC01Essential Chilled Water Valve Verification13

43ST-2EC01Essential Chilled Water Valve Verification9

55OP-0GT01Gas Turbine Generator #1 Operating Instructions 46

55OP-0GT02Gas Turbine Generator #2 Operating Instructions44

70DP-0MR01Maintenance Rule14

70DP-9GT01Gas Turbine Generator (GTG) Test RecordSeptember 22,199370TI-9ZC01Boric Acid Walkdown Leak Detection 6

73DP-9EE02Inservice Inspection Examination Activities8

73DP-9XI03ASME Section XI Inservice Inspection6

73ST-9AF02AFA-P01 Inservice Test36

73ST-9EC01Essential Chilled Water Pumps - Inservice Test16

Procedures:NUMBERTITLEREVISION /DATEAttachmentA-2873ST-9SG01MSIVs - Inservice Test2673ST-9SG05ADV Nitrogen Accumulator Drop Test23

73ST-9XI20Atmosphere Dump Valves (ADV)- Inservice Test 20

73ST-9ZZ20ASME Section XI Off-Line Set Pressure Verification22

74DP-9CY04System Chemistry Specifications46

74DP-9DF01Diesel Fuel Oil Program5

74ST-9DF02Diesel Generator Fuel Oil Receipt Surveillance Test4

81DP-0EE10Plant Modifications 12

90DP-0IP10Condition Reporting32

A0-104-W311-104Tank Elevation

FSN-T02 1A0-W-FSP-300Water Reclamation Plant Fuel Oil System P&ID Rev 11.Fuel Oil SampleAPS Water Reclamation Facility- Service Report FuelTank Thermal Stability.June 28, 2001GTG1.02Gas Turbine Generator (GTG) Test Record October 20,1993Sample 10923Fuel Oil Sample Data for Gas Turbines June 28, 2001

WROP- 8FS01WRF Fuel System (FS) Operating Procedure7Scenarios and Job Performance Measures (JPMs)NUMBERTITLEScenarioFailure of Spray Pond pump to start/Align Class AF pump suctions tothe Refueling Makeup Water Tank (RMWT)

AttachmentA-29ScenarioGas Turbine Generator (GTG) FailureScenarioAlign Auxiliary Feedwater Pump "N" (AFN) suction manually ScenarioFailure of Containment Spray Pump/Class Battery Charger Alignment ScenarioFeed with Condensate/Failure to depressurizeJPMIn-plant performance of 40OP-9PK01 section 4.5 to place the ACbattery charger on

PKA-M41JPMIn-plant performance to manually open CTA-HV-1

JPMPlant Performance of GTG Start

JPMIn-Plant Performance of Attachment 80A

JPMIn-plant Performance of Alarm Response to Align

AFB-P01 to theRMWSurveillance Test Work Order Results:NUMBERTITLEDATE2662494Auxiliary Feedwater System Surveillance Test Results January 18, 2005

2663158Auxiliary Feedwater System Surveillance Test Results January 25, 2005

2663156Auxiliary Feedwater System Surveillance Test Results February 15, 2005

2662788Auxiliary Feedwater System Surveillance Test Results February 21, 20052662882Auxiliary Feedwater System Surveillance Test Results March 17, 20052662496Auxiliary Feedwater System Surveillance Test Results March 22, 2005

2696235Auxiliary Feedwater System Surveillance Test Results April 12, 2005

2696505Auxiliary Feedwater System Surveillance Test Results April 18, 2005

2751270Auxiliary Feedwater System Surveillance Test Results May 5, 2005

2665891Auxiliary Feedwater System Surveillance Test Results June 9, 2005

Surveillance Test Work Order Results:NUMBERTITLEDATEAttachmentA-302696237Auxiliary Feedwater System Surveillance Test Results June 15, 20052701800Auxiliary Feedwater System Surveillance Test Results July 5, 2005

2703435Auxiliary Feedwater System Surveillance Test Results July 15, 2005

2703430Auxiliary Feedwater System Surveillance Test Results August 1, 2005

2702537Auxiliary Feedwater System Surveillance Test Results August 9, 2005

2702932Auxiliary Feedwater System Surveillance Test Results August 31, 2005

2701805Auxiliary Feedwater System Surveillance Test Results September 9, 2005

2701801Auxiliary Feedwater System Surveillance Test Results September 27, 2005

2703436Auxiliary Feedwater System Surveillance Test Results October 3, 2005

2703431Auxiliary Feedwater System Surveillance Test Results October 24, 2005

2702538Auxiliary Feedwater System Surveillance Test Results November 2, 2005

2702933Auxiliary Feedwater System Surveillance Test Results November 22, 2005

2701806Auxiliary Feedwater System Surveillance Test Results December 11, 2005

2767668Auxiliary Feedwater System Surveillance Test Results December 14, 2005

2703437Auxiliary Feedwater System Surveillance Test Results December 27, 2005

27996Auxiliary Feedwater System Surveillance Test Results January 17, 2006

27905Auxiliary Feedwater System Surveillance Test Results January 25, 2006

27627Auxiliary Feedwater System Surveillance Test Results February 14, 2006

27206Auxiliary Feedwater System Surveillance Test Results February 24, 2006

27203Auxiliary Feedwater System Surveillance Test Results March 15, 2006

Surveillance Test Work Order Results:NUMBERTITLEDATEAttachmentA-312728006Auxiliary Feedwater System Surveillance Test Results March 21, 20062727906Auxiliary Feedwater System Surveillance Test Results April 17, 2006

27998Auxiliary Feedwater System Surveillance Test Results May 4, 2006

2819601Auxiliary Feedwater System Surveillance Test Results May 5, 2006

27629Auxiliary Feedwater System Surveillance Test Results May 11, 2006

27207Auxiliary Feedwater System Surveillance Test Results May 27, 2006

2768430Auxiliary Feedwater System Surveillance Test Results June 6, 2006

2899504Auxiliary Feedwater System Surveillance Test Results June 11, 2006

2769608Auxiliary Feedwater System Surveillance Test Results June 13, 2006

2768770Auxiliary Feedwater System Surveillance Test Results June 26, 2006

2769606Auxiliary Feedwater System Surveillance Test Results July 5, 2006

2768772Auxiliary Feedwater System Surveillance Test Results July 10, 2006

2902608Auxiliary Feedwater System Surveillance Test Results July 20, 2006

2768824Auxiliary Feedwater System Surveillance Test Results August 3, 2006

2768824Auxiliary Feedwater System Surveillance Re-TestResults August 3, 20062768433Auxiliary Feedwater System Surveillance Test Results August 7, 20062769607Auxiliary Feedwater System Surveillance Test Results August 21, 2006

2768431Auxiliary Feedwater System Surveillance Test Results August 29, 2006

2769609Auxiliary Feedwater System Surveillance Test Results September 8, 2006

2793600Auxiliary Feedwater System Surveillance Test Results September 25, 2006

AttachmentA-32Vendor Manuals:NUMBERTITLEREVISION13-VTD-C628-00051Cooper Energy Instruction Manual for KSV Turbocharged Diesel Generating Unit for Nuclear Power Plant Emergency Stand-by Service (Pub. # 010997)1013-VTD-S903-00002Solar Turbines Systems Operator's Guide for CentaurTaurus Gas Turbine-Driven Generator Set

(Pub. #

SOG-93-45521)3SM-100, Section 3.2Ideal Electric Instruction Manual for SynchronousMotors, Generators, D.C. Exciters & Brushless EquipmentWork Orders:NUMBERTITLEDATE2565413Inspect/Lubricate and Overhaul of G.E./VAC VacuumCircuit BreakersOctober 25,20032611733Inspect/Lubricate and Overhaul of G.E./VAC VacuumCircuit BreakersApril 24, 20042647003Inspect/Lubricate and Overhaul of G.E./VAC VacuumCircuit BreakersOctober 11,20042868734Inspect AENANX02 Startup TransformerJanuary 9,20072868735Inspect/Test Transformer, Perform Procedure 32MT-9NA03 and Cycle the Tie-BreakersFebruary 27,20072868839Obtain AENANX02 Transformer Oil Sample While theTransformer is EnergizedJanuary 12,20072869536Calibrate the S/U XFMR Instruments and Perform theLarge Transformer Internal CT TestingFebruary 14,20072890298Predictive Maintenance Group to Perform Thermographyper 37TI-9ZZ01January 30,2007

Work Orders:NUMBERTITLEDATEAttachmentA-332914801GTG #1 Control Battery Failed the Discharge Test on7/5/06, Need to Re-perform the Capacity Discharge Test per DFWO DispositionMarch 20,20072914802GTG #1 Diesel Start Battery Failed the Discharge Teston 7/6/06, Need to Re-perform the Capacity Discharge Test per DFWO Disposition March 20,2007