IR 05000275/2006011

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IR 05000275-06-011; 05000323-06-011; September 11 Through October 6, 2006; Diablo Canyon Nuclear Power Plant, Units 1 and 2: NRC Inspection Procedure 71111.21, Component Design Basis Inspection
ML070560003
Person / Time
Site: Diablo Canyon  Pacific Gas & Electric icon.png
Issue date: 02/23/2007
From: Jones W B
NRC Region 1
To: Keenan J S
Pacific Gas & Electric Co
References
IR-06-011
Download: ML070560003 (24)
v • d • e


Text

February 23, 2007

John S. KeenanSenior Vice President - Generation and Chief Nuclear Officer Pacific Gas and Electric Company P.O. Box 770000 Mail Code B32 San Francisco, CA 94177-0001

SUBJECT: DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2, NRCCOMPONENT DESIGN BASIS INSPECTION REPORT 05000275/2006011; 05000323/2006011

Dear Mr. Keenan:

On January 11, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an inspectionat your Diablo Canyon Nuclear Power Plant, Units 1 and 2. The preliminary findings were discussed on October 6, 2006, with you and other members of your staff. After additional in-office inspection, a final telephonic exit meeting was conducted on January 11, 2007, with Ms. D. Jacobs, Vice President, Nuclear Services, and other members of your staff. The enclosed report documents the inspection findings.The team examined activities conducted under your license as they relate to safety andcompliance with the Commission's rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel. Based on the results of this inspection, the NRC has identified two issues that were evaluatedunder the risk significance determination process as having very low safety significance (green). The NRC has also determined that violations are associated with these issues. These violations are being treated as noncited violations, consistent with Section VI.A of the Enforcement Policy. These noncited violations are described in the subject inspection report. If you contest the violation 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 Diablo Canyon Nuclear Power Plant, Units 1 and 2.

Pacific Gas and Electric Company-2-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: 50-275; 50-323Licenses: DPR-80; DPR-82

Enclosures:

Inspection Report 05000275/2006011; 05000323/2006011 w/Attachment Supplemental Informationcc w/enclosures:

Donna Jacobs Vice President, Nuclear Services Diablo Canyon Power Plant P.O. Box 56 Avila Beach, CA 93424James R. Becker, Vice President Diablo Canyon Operations and Station Director, Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 56 Avila Beach, CA 93424Sierra Club San Lucia ChapterATTN: Andrew Christie P.O. Box 15755 San Luis Obispo, CA 93406Nancy CulverSan Luis Obispo Mothers for Peace P.O. Box 164 Pismo Beach, CA 93448 Pacific Gas and Electric Company-3-ChairmanSan Luis Obispo County Board of Supervisors County Government Building 1055 Monterey Street, Suite D430 San Luis Obispo, CA 93408Truman Burns\Robert KinosianCalifornia Public Utilities Commission 505 Van Ness Ave., Rm. 4102 San Francisco, CA 94102-3298Diablo Canyon Independent Safety CommitteeRobert R. Wellington, Esq.

Legal Counsel 857 Cass Street, Suite D Monterey, CA 93940Director, Radiological Health BranchState Department of Health Services P.O. Box 997414 (MS 7610)

Sacramento, CA 95899-7414Richard F. Locke, Esq.Pacific Gas and Electric Company P.O. Box 7442 San Francisco, CA 94120City EditorThe Tribune 3825 South Higuera Street P.O. Box 112 San Luis Obispo, CA 93406-0112James D. Boyd, CommissionerCalifornia Energy Commission 1516 Ninth Street (MS 34)

Sacramento, CA 95814Jennifer TangField Representative United States Senator Barbara Boxer 1700 Montgomery Street, Suite 240 San Francisco, CA 94111 Pacific Gas and Electric Company-4-Electronic distribution by RIV:Regional Administrator (BSM1)DRP Director (ATH)DRS Director (DDC)DRS Deputy Director (RJC1)Senior Resident Inspector (TWJ)Branch Chief, DRP/B (GEW)Senior Project Engineer, DRP/E (FLB2)Team Leader, DRP/TSS (RLN1)RITS Coordinator (KEG)DRS STA (DAP)V. Dricks, PAO (VLD)J. Lamb, OEDO RIV Coordinator (JGL1)ROPreports DC Site Secretary (AWC1)SUNSI Review Completed: __Y___ADAMS: Yes G No Initials: _WCS_____ Publicly Available G Non-Publicly Available G Sensitive Non-SensitiveSRI:EB1SOE:OBRI:EB1RI:EB1C:EB1C:PBBC:EB1WCSifre/lmbRELantzJHNadelJPReynosoWBJonesVGaddyWBJones/RA//RA//RA//RA//RA//RA//RA/2/15/072/20/072/21/072/21/072/21/072/22/072/23/07OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax Enclosure-1-ENCLOSUREU.S. NUCLEAR REGULATORY COMMISSION REGION IV Dockets:50-275; 50-323 Licenses:DPR-80; DPR-82 Report No.:05000275/2006011; 05000323/2006011 Licensee:Pacific Gas and Electric Company Facility:Diablo Canyon Nuclear Power Plant, Units 1 and 2 Location:7 1/2 miles NW of Avila Beach Avila Beach, California Dates:September 11, 2006, through January 11, 2007 Team Leader:W. Sifre, Senior Reactor Inspector, Engineering Branch 1 Inspectors:P. Gage, Senior Operations EngineerR. Lantz, Senior Emergency Preparedness Inspector J. Nadel, Reactor Inspector, Engineering Branch 1 J. Reynoso, Reactor Inspector, Engineering Branch 1Contractors:F. Baxter, Electrical, Beckman and AssociatesM. Shlyamberg, Mechanical, Nuenergy, Inc.Accompanied By:A. Fairbanks, Reactor Inspector (NSPDP)M. Young, Reactor Inspector (NSPDP)Approved By:William B. Jones, ChiefEngineering Branch 1 Division of Reactor Safety Enclosure-2-

SUMMARY OF FINDINGS

IR 05000275/2006011; 05000-323/2006011; September 11 through October 6, 2006; DiabloCanyon Nuclear Power Plant, Units 1 and 2: NRC Inspection Procedure 71111.21, ComponentDesign Basis Inspection

.The report covered a period of inspection by a team of seven inspectors and two contractors. Two findings of very low safety significance 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 processdoes 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: Initiating Events

Green.

The team identified a noncited violation of 10 CFR Part 50, Appendix B,Criterion III, Design Control, for the failure to translate design basis information intospecifications and procedures. The team identified that a nonconservative flow rate was used as an input in engineering design calculations resulting in the potential for choked flow at the discharge valves for the Unit 1 auxiliary service water system. Choked flow turbulence is a wear concern for these components, and can result in system failure.

The licensee entered this finding into their corrective action program as Action Requests A0678338 and A0678472.The finding is more than minor because the error affected the Mitigating SystemCornerstone objective (Design Control attribute) of ensuring availability, reliability, and capability of the auxiliary service water systems to respond to initiating events to prevent undesired consequences. Using the Manual Chapter 0609, Significance DeterminationProcess, Phase 1 screening worksheet, the issue screened as having very low safetysignificance because 1) did not represent a loss of system safety function; and 2) did not represent an actual loss of safety function of one or more non-technical specification trains of equipment; and did not screen as potentially risk significant because of a seismic, flooding, or sever weather initiating event (Section 1R21b.1.).*Green. The team identified a noncited violation of 10 CFR Part 50, Appendix B,Criterion III, Design Control, for the failure to demonstrate that the acceptance criteriafor surveillance tests had conservatively accounted for uncertainties in determination of the minimum allowed ultimate heat sink temperature. Specifically, the team identified that the acceptance criteria specified in the Surveillance Test Procedure STP I-1A,

Routine Shift Checks Required by the Licensee, Revision 101, did not correctly accountfor instrument uncertainty. The licensee entered this finding into their corrective action program as Action Request A0682398.

Enclosure-3-The finding is more than minor because the error affected the Mitigating Systemcornerstone objective (Design Control attribute) of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Using the Manual Chapter 0609, Significance Determination Process

,Phase 1 screening worksheet, the issue screened as having very low safety significance because 1) did not represent a loss of system safety function; and 2) did not represent an actual loss of safety function of one or more non-technical specification trains of equipment; and did not screen as potentially risk significant because of a seismic, flooding, or severe weather initiating event (Section 1R21b.2).B.Licensee-Identified Findings None.

Enclosure-4-

REPORT DETAILS

1.REACTOR SAFETYCornerstones: Initiating Events/Mitigating Systems/Barrier Integrity1REACTOR 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.

1R21 Component Design Bases Inspection (71111.21)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' analysis methods. 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 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 because of modification, or margin reductions identified as a result of material condition issues. Equipment reliability issues were alsoconsidered 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

-5-resident inspector input of problem equipment; system health reports; industry operatingexperience; and licensee 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 lowdesign margin components, three to five relatively high-risk operator actions, and 4 to 6 operating experience issues. The sample selection for this inspection was 18 components, 8 operator actions, and 6 operating experience items. The components selected for review were:

  • Component cooling water surge tank*Component cooling water motor-operated valves
  • Component cooling water piping integrity
  • Component cooling water heat exchangers
  • Switchgear ventilation (fans, louvers, etc.), 480v, (Fans 43,44)
  • 125vdc batteries (capacity, fuses)
  • Emergency diesel generators (engines, output breakers)
  • Battery chargers (electrolytic capacitors)
  • Main steam isolation valves
  • Auxiliary salt water pumps
  • Auxiliary salt water motor-operated valves
  • Power operated relief valves
  • Residual heat removal pump
  • Residual heat removal containment isolation valvesThe selected operator actions were:
  • Response to a spurious safety injection actuation at power
  • Transfer to cold leg recirculation with failure of one residual heat removal pumpto trip*Transfer to hot leg recirculation
  • Isolation of design leakage into the component cooling water system
  • Reduction of component cooling water heat loads because of a malfunction inthe component cooling water system, establishment of an alternate cooling source for the component cooling water system*Diagnosis and response to a loss of secondary heat sink

-6-*Establishment of alternate cooling to the 480 Volt AC switchgear*Restoration of auxiliary saltwater cooling via the unit auxiliary salt water systemcross-connect valve (FCV-601) during a station blackoutThe operating experience issues were:

  • Auxiliary feedwater pump recirculation line orifice fouling - potential commoncause failure at Point Beach*Sizing and setting of molded case circuit breakers and thermal overload heaters
  • Vibration induced degradation of butterfly valves (Fisher valves)
  • Guidance on developing acceptable inservice testing programs
  • Periodic verification of design-basis capability of safety-related motor-operatedvalves

b. Findings

b.1.Failure to Use Correct Design Inputs in Determination of a Potential for ChokingFlow/Cavitation Across the Auxiliary Service Water Throttled Butterfly ValvesIntroduction. The team identified a Green noncited violation of 10 CFR Part 50,Appendix B, Critieria III, Design Control, for the failure to translate design basisinformation into specifications and procedures. Specifically, the team identified that a non-conservative flow rate was used as an input in engineering design calculations (M-988, ASW System Flows, Pressures and Temperatures, Revision 6 and SurveillanceTest Procedure 1&2STP-M 26, ASW System Flow Monitoring, Revision 25B and M-885,Determine ASW System Flow in Various ASW/CCW Configurations and Conditions

,Revision 3).

Description.

The team identified that conditions for choked flow in the auxiliary saltwater system are established when the auxiliary salt water system is aligned in a two pump and one heat exchanger configuration. Since mid-1990 the auxiliary salt water trains have been aligned in this configuration when the ultimate heat sink temperature was in excess of 64°F and under an accident condition because of the normally open cross-train connecting Valves FCV-495 and FCV-496. The team determined that the auxiliary salt water system maximum auxiliary salt water system flow rate across the component cooling water heat exchanger outlet throttled butterfly valves following a design basis event would be significantly higher than the licensee assumed in Calculation M-988. In response to the team's questions the licensee issued Action Requests A0678338, Calculation M-988 Discrepancies, September 26, 2006; andA0678472, Incipient Cavitation of ASW Operation w/2PP, 1HX, September 27, 2006.

-7-Action Request A0678472 documented that Calculation M-988 used a non-conservativeflow rate for the cavitation determination and changed Unit 1 valves from 55 degrees open to greater than 70 degrees open (Unit 2 valve positions were greater than 70 degrees open). The licensee provided the following justification for the operability of the valves under choking conditions. "With 1 minute of 2 pumps, one heat exchanger operation per swap, and 50 swaps per year, we have accumulated at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of operation in the alignment of concern over the last decade. The maintenance history of these valves does not indicate any unusual damage or parts replaced because of cavitation damage. In addition, these valves are inspected when the heat exchangers are opened for cleaning every outage and no signs of damage or degradation that would interfere with valve operation have been noted." The team determined that the evidence of no "unusual damage or parts replaced because of cavitation damage" was based on a system operation under normal conditions where the auxiliary service water temperature is rarely above 70F. Under accident conditions for which the choked flowwas predicted, the auxiliary service water temperature was expected to be at least

140F. An elevated temperature could significantly increase the cavitation potential. Also given the transient nature of the swap over, the time period is not sufficiently long enough to evaluate for a cumulative effect. Based on the team's review of this action request, the licensee revised the operability determination and issued Action Request A0679175, Review Performance of Operability Determinations During CDBI, October 5, 2006.Analysis. The failure to use a conservative design input in the engineering analysis wasa performance deficiency. This issue is more than minor because the error affected the Mitigating System cornerstone objective (Design Control attribute) of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Using the Manual Chapter 0609, Phase 1 screening worksheet, the issue screened as having very low safety significance because it did not represent a loss of system safety function and it did not represent an actual loss of safety function of one or more non-technical specification trains of equipment; and it did not screen as potentially risk significant because of a seismic, flooding, or severe weather initiating event.Enforcement. Part 50 of Title 10 of the Code of Federal Regulations, Appendix B,Criterion III, Design Control, states, in part, that measures shall be established to assurethat design basis are correctly translated into specifications and procedures. Contraryto the above, in Calculation M-988, the licensee did not use a conservative auxiliary service water flow rate to determine the outlet valve position necessary to prevent choke flow and subsequent component damage or even a catastrophic failure of the valves during a design basis event.Because the finding is of very low safety significance (Green) and has been entered intothe licensee's corrective action program (Action Requests A0678338 and A0678472), it is a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy:

NCV 05000275; 323/2006011-01, Failure to Use Correct Design Input in Determination of a Potential for Chocking Flow/Cavitation Across the Auxiliary Service Water Throttled Butterfly Valves.

-8- b.2.Failure to Consider Instrument Uncertainty in Surveillance Requirements for TechnicalSpecifications LCO 3.7.9Introduction. A noncited violation of very low safety significance (Green) was identifiedfor the failure to demonstrate that the acceptance criteria for surveillance tests had conservatively accounted for uncertainties in the determination of the minimum allowed ultimate heat sink temperature. Specifically, the team identified that the acceptance criteria specified in the Surveillance Test Procedure STP I-1A, Routine Shift ChecksRequired by the Licensee, Revision 101, did not correctly account for the instrumentuncertainty.Description. Technical Specifications LCO 3.7.9 requires placing the second closedloop cooling system heat exchanger in service if the ultimate heat sink temperature exceeds 64F. The team's review of Calculation Support STP I-1A, IndicatedTemperature Uncertainties for TI-311 /-328, Revision 0, identified that the instrumentcorrection for the maximum ultimate heat sink temperatures specified in this calculation and in the Surveillance Test Procedure STP I-1A (SR 3.7.9.2) were non-conservative.

The licensee's correction for the instrument uncertainty value was based on a standard deviation and not 2.4F, the actual uncertainty of an individual instrument loop. Thelicensee used an instrument uncertainty correction of 0.9F when two instruments wereavailable and 0.4F when three or four instruments were available. Use of standarddeviation methodology is not an NRC and industry recognized methodology and was not in accordance with the licensee's programs and procedures for setpoint uncertainties (CF6.ID1, Setpoint Control Program; CF6.NE1, Instrument ChannelUncertainty and Setpoint Methodologies; AWP E 001, Development of PME ChannelUncertainty Calculations). Furthermore, use of standard deviation in this case is notscientifically valid, since this method accounts for only the statistical performance of the individual loops and cannot be used because of the variation in the monitored variable -

temperatures of the separate temperature channels. Therefore, the surveillance correction factors were non-conservative. The licensee issued Action Request A0682398, Evaluate UHS Instrument Uncertainty Used in STP I-1A

,November 14, 2006. This action request stated that the STP acceptance criteria will be revised to a value that is based on the approved methodologies.Analysis. The failure to use a conservatively determined instrument uncertainty in thederivation of the acceptance criteria for the technical specifications surveillance values was a performance deficiency. This issue is more than minor because the error affected the Mitigating System cornerstone objective (Design Control attribute) of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Using the Manual Chapter 0609,Phase 1 screening worksheet, the issue screened as having very low safety significance because it did not represent a loss of system safety function and it did not represent an actual loss of safety function of one or more non-technical specification trains of equipment; and did not screen as potentially risk significant because of a seismic, flooding, or severe weather initiating event.

-9-Enforcement. Part 50 of Title 10 of the Code of Federal Regulations, Appendix B,Criterion III, states, in part, that measures shall be established to assure that designbasis are correctly translated into specifications and procedures. Contrary to the above, the licensee did not conservatively account for the effect of the instrument uncertainty in derivation of the acceptance criteria for the technical specifications surveillance values for Technical Specification LCO 3.7.9; thus, the maximum allowable ultimate heat sink temperature for a single component cooling water heat exchanger operation could not be guaranteed at the Technical Specification LCO values.Because the finding is of very low safety significance (Green) and has been entered intothe licensee's corrective action program (Action Request A0682398), it is a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000275/323/2006011-02, Failure to Consider Instrument Uncertainty in Surveillance Requirements for Technical Specifications LCO

OTHER ACTIVITIES

4OA6Meetings, Including ExitOn January 11, 2007, the team leader presented the inspection results, via telephone,to, Ms. D. Jacobs, Vice President, Nuclear Services and other members of the Diablo Canyon Power Plant's staff who acknowledged the findings. The inspectors confirmed that proprietary information was provided and examined during this inspection.ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

T. Baldwin, Supervisor, Engineering
J. Ballard, Engineer
T. Chitwood, Senior Operations Engineer
C. Dougherty, Senior Engineer, Regulatory Services
J. Fields, Auditor, Quality Verification
L. Fuseo, Manager, Engineering Services
C. Harbor, Manager, Performance Improvement
D. Jacobs, Vice President, Nuclear Services
R. Klimezak, Manager, Engineering Services
M. Mayer, Engineering Supervisor
P. Nugent, Manager, Project Engineering
L. Parker, Supervisor, Regulatory Services
P. Roller, Director, Performance Improvement
K. Schrader, Engineer, Regulatory Services
R. Washington, Instrumentation and Controls Supervisor
R. West, Engineering Supervisor
S. Westcott, Manager, Engineering Services
S. Zawalick, Engineer, Regulatory Services

NRC personnel

T. Jackson, Senior Resident inspector
T. Brown, Resident Inspector
T. McConnell, Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and

Closed

05000275; 323/2006011-01NCVFailure to Use Correct Design Inputs inDetermination of a Potential for Choking
Flow/Cavitation Across the Auxiliary Service
Water Throttled Butterfly Valves05000275; 323/2006011-02NCVFailure to Consider Instrument Uncertainty inSurveillance Requirements for Technical
Specifications LCO 3.7.9
AttachmentA-2

LIST OF DOCUMENTS REVIEWED

CalculationsNumberTitleRevision/DateN-013MOV Limiting Process Condition Evaluation (GL 89-10)5N-124System Check Valve Safety Function Design Basis3DC2-SJ-50711Steam Generator PORV Setting change 0N-791Maximum Steam Generator Pressure Following a PlantTrip 1M69-00606Air Cylinder Sizing MSIV Unit 1 and 2January,1977M-1019Heat Removal Capability of the CCW / RHR Following aLOCA 0M-1017CCW Flow Balancing3WCAP 14825Unit 1 Power Update Impact to Balance of Plant0

WCAP 14282Evaluation of Peak CCW Temperature Scenarios forDiablo Canyon
1M-988Evaluation of Effects of New Auxiliary Service Water Bypass Piping
6M-938CCW Data Input Fro 1993 Containment AnalysisProgram 1M-305CCW System Temperature and Pressure U-120M-966Maximum CCW Flow Rate to the RHR Heat ExchangersApril,1994099-DC4.16 kV High Resistance Grounding System4182B-DC4.16 kV High Resistance Grounding System4M-786Determine the Required Diesel Fuel Oil Storage Neededto Meet Licensing Basis for Operating Minimum ESF
Loads 14277-DCEmergency Diesel Generator Opposite Unit Operation0359-DCDetermination of 230 kV Grid Capability Limits as aDCPP Power Source
8359B-DCEvaluation of Voltage Transients on ESF LoadSequencing
0366A-DCSizing & Setting of Molded Case Circuit Breakers &Thermal Overload Heaters
CalculationsNumberTitleRevision/DateAttachmentA-3235E-DCBattery 22 Sizing, Load Flow, Voltage Drop, ShortCircuit & Charger Sizing Calculation
235B-DCBattery 12 Sizing, Load Flow, Voltage Drop, ShortCircuit & Charger Sizing Calculation
8357A-DCUnits 1 & 2 Load Flow, Short Circuit, and Motor Starting12357E-DCGuidelines of Transformer Data Entered Into ETAP683-46Determine the Adequacy of Natural Ventilation toMaintain Hydrogen Concentration Inside the Battery Rooms Within Allowable Concentration Limit.
5J-103ADetermine Channel Uncertainty for Auxiliary ServiceWater Flowmeters
FIT-484 and FIT-485
0J-139Uncertainty in Using the AMAG Ultrasonic Flowmeter forCalibration Checks on the Auxiliary Service Water Auxiliary Service Water
Magmaster Magnetic Flowmeters
0M-384AFW, Restricting Orifice for Turbine Bearing & GovernorCoolers 2M-885Determine Auxiliary Service Water
System Flow inVarious Auxiliary Service Water /CCW Configurations and Conditions
3M-988Auxiliary Service Water System Flows, Pressures andTemperatures
6M-1100Surveillance of Condensate Storage Tanks and FireWater Tank to Ensure Compliance with Technical Specification 3.7.6
0STA-243Assessment of the Potential for Film Boiling in the CCWHX Tubes 0STA-244RELAP5 Evaluation of Auxiliary Service Water/CCW HX
Subcooling
0WCAP-14282Evaluation of Peak CCW Temperature Scenarios forDiablo Canyon Units 1 and 2
CalculationsNumberTitleRevision/DateAttachmentA-4N-124Safety Related Check Valves in Various Systems4G-026-01 Study
2086CCW Design I Piping Analysis2PME-14-PI-2030CCW Surge Tank Pressure Indications (PI-2030/2031and P0710A) Computer Point
0PME-14-PI-2030CCW Surge Tank Pressure Indications (PI-2030/2031and P0710A) Computer Point
1AR PK01-14CCW Surge Tank Pressure4AAR PK01-14CCW Surge Tank Pressure5
AR PK01-14CCW Surge Tank Pressure8A
AR PK01-14CCW Surge Tank Pressure9Action Requests
A0053753 A0053757
A0055760
A0055764
A0062550
A0104405
A0104406
A0140107
A0165357
A0265248
A0278874
A0309356
A0327420
A0340549
A0384028
A0412152
A0497372
A0504788
A0505145
A0524592
A0524592
A0531252
A0539241
A0544744
A0545206
A0552440
A0558344 A0560825 A0561002
A0563111
A0563407
A0563925
A0565571
A0572553
A0574410
A0574410
A0575264
A0575290
A0593352
A0594216
A0595213
A0598307
A0601794
A0604521
A0604708
A0605055
A0605108
A0607758
A0615021
A0615938
A0616630
A0616828
A0618343
A0618344 A0618728 A0618728
A0618730
A0619640
A0625440
A0629284
A0629346
A0630143
A0630181
A0630182
A0631050
A0631353
A0632886
AO634661
A0635344
A0635441
A0636885
A0637723
A0638509
A0643107
A0643211
A0643434
A0644020
A0645680
A0645716
A0645846 A0646263 A0647194
A0648857
A0649405
A0649741
A0649921
A0650143
A0650302
A0652663
A0652972
A0653325
A0656726
A0658049
A0661523
A0662539
A0663619
A0664426
A0665260
A0665707
A0665753
A0665756
A0666177
A0666993
A0667739
A0668073
A0669758 A0670430 A0670462
A0670789
A0670796
A0672304
A0672424
A0672590
A0672862
A0672863
A0673550
A0675689
A0675762
A0676103
A0676422
AO677521
A0677581
A0677581
AO677709
AO677709
AO677711
A0677716
A0677849
A0677909
AO678002
A0678029
A0678132 A0678280 A0678338
A0678411
A0678432
A0678443
A0678472
A0678537
A0678541
A0678542
A0678642
A0678650
A0678666
A0678768
A0678787
A0678821
A0678931
A0678932
A0678949
A0678980
A0678986
A0679115
A0679140
A0679175
A0679198
A0682398
A0682549
AttachmentA-5DrawingsNumberTittleRevision/Date102017, Sheet 1Unit 1, Piping Schematic Saltwater Systems89102017, Sheet 3Unit 1, Piping Schematic Screen Wash Systems101
2017, Sheet 3BUnit 1, Piping Schematic Saltwater System92
2014, Sheet 1Unit 1, Piping Schematic Component CoolingWater System
60102014, Sheet 5Unit 1, Piping Schematic Component CoolingWater System
57102014, Sheet 5AUnit 1, Piping Schematic Component CoolingWater System
45102014, Sheet 10Unit 1, Piping Schematic Component CoolingWater System
60DC663316Mission Valve Schematic6106714 Sheet 2CCW System56
500989Drainage and Fire Fighting Turbine Building at85' Elevation
9500116Drainage and Fire Fighting Turbine Building at85' Elevation
13107717CCW System Drawing, Sheet 153107717CCW System Drawing, Sheet 250
107717CCW System Drawing, Sheet 344
107717CCW System Drawing, Sheet 451
107717CCW System Drawing, Sheet 517
107717CCW System Drawing, Sheet 646
107717CCW System Drawing, Sheet 752
107717CCW System Drawing, Sheet 849
107717CCW System Drawing, Sheet 949
2028CCW System ASME Section XI PipingSchematic, Sheet 2
21102028CCW System ASME Section XI PipingSchematic, Sheet 25
DrawingsNumberTittleRevision/DateAttachmentA-6102028CCW System ASME Section XI PipingSchematic, Sheet 26
11102028CCW System ASME Section XI PipingSchematic, Sheet 27
48102028CCW System ASME Section XI PipingSchematic, Sheet 28
17102028CCW System ASME Section XI PipingSchematic, Sheet 29
34102028CCW System ASME Section XI PipingSchematic, Sheet 30
31102028CCW System ASME Section XI PipingSchematic, Sheet 31
54102028CCW System ASME Section XI PipingSchematic, Sheet 32
2DC
663049 Assembly of MSIV13DC 663217-110-1Tube Sheet Bundle, RHRMarch 10, 1969
DC 663217-112-1RHR Shell and Channel DetailsMarch 12. 1969
DC 663217-113-1RHR Tube Supports at Tube BendsSeptember 27, 1972
DC 663217-111-1RHR Tube Supports at Tube BendsNovember 27, 1972
DC-663212-26-1CCW Heat Exchangers Heat Transfer Curves August, 1970
PGE-94-662Heat Transfer Coefficients for the RHR HeatExchangersOctober, 6, 1994PGE-00-503RHR Heat Exchanger Flow Increase EvaluationFebruary 8, 2000DC 663217-15-2RHR System Description1
DC 663217-117-1Increase in Design Temperature for the shellside of the Diablo Canyon RHR heat exchangers
0502110Single Line Dgm., 500/230/25/12/4.16 kVSystems 15437533Single Line Meter & Relay Dgm., 4160 V System35441228Single Line Meter & Relay Dgm., 4160 VSystem, Bus Section D & E
16441230Single Line Meter & Relay Dgm., 4160 VSystem, Bus Section G & H
DrawingsNumberTittleRevision/DateAttachmentA-7441229Single Line Meter & Relay Dgm., 4160 VSystem, Bus Section F
16437916Single Line Meter & Relay Dgm., 480 V System,Bus Section 1F
2441237Single Line Meter & Relay Dgm., 480 V System,Bus Section 2F
30437542Single Line Meter & Relay Dgm., 480 V System,Bus Section 1G
47441238Single Line Meter & Relay Dgm., 480 V System,Bus Section 2G
41437543Single Line Meter & Relay Dgm., 480 V System,Bus Section 1H
43441239Single Line Meter & Relay Dgm., 480 V System,Bus Section 2H
38441240Single Line Meter & Relay Dgm., 125 V DCSystem 33445295Single Line Meter & Relay Dgm., 125 V & 250 VDC System 10437545Single Line Diagram Pressurized Heaters10437518Single Line Diagram for Station Auxiliaries38
441220Single Line Diagram for Station Auxiliaries25
494432Auxiliary Building Switchgear Room Supply FanS-43 & S-44
2445086Schematic Diagram Ventilating Fan Motors4437589Schematic Diagram SI Pumps6
437665Schematic Diagram 4 kV Diesel Generators andAssociated Circuit Breakers
24102036PORV Valves, Sheet 7F106441306Schematic Diagram Reactor Coolant MOVs16
437587Schematic Diagram Reactor Coolant MOVs20
106714Component Cooling Water - Unit 156
457395Tornado Barrier for Diesel Engine GeneratorRoom Turbine Building, Area "A"
AttachmentA-8 ProceduresNumberTittleRevision1 STP P-ASW-A11Routine Surveillance Test of Auxiliary Saltwater Pump
1-1231&2 STP P-ASW-APerformance Test of Auxiliary Saltwater Pumps22
1&2 STP M-26auxiliary service water
System Flow Monitoring25B
OP D-1:VAuxiliary Feedwater System - Alternate AuxiliaryFeedwater Supplies

& 18STP I-1ARoutine Shift Checks Required by the Licensee101 &

2FR-H.1Response to Loss of Secondary Heat Sink21
F-0Critical Safety Function Status Trees13A
E-0Reactor Trip or Safety Injection30
E-1Loss of Reactor or Secondary Coolant22
E-1.4Transfer to Hot Leg Recirculation16
E-1.1Safety Injection Termination21
G-1Emergency Classification and Emergency Plan Activation34
PEP
EN-1Plant Accident Mitigation Diagnostic "Aids and Guidelines15
OP-AP-11Malfunction of the Component Cooling Water System,Appendix B, CCW Heat Load Isolation
21AR-PK-15-19ESF 480V/DC RMS Vent Trouble8AR-PK-15-06Control Room Vent17
AR-PK-15-09Electrical Rooms Temp Monitor27
AR-PK-10-15Fire Alarm Trouble7
AR-PK-11-21High Radiation25A
AR-PK-01-07CCW System Surge Tk LVL/MK-UP13
OP H-10:1Auxiliary Building Switchgear Ventilation - Make Availableand System Operation
27OP1.DC11Attachment 7.2, Time-limiting Operator Actions02/07/06OP A-2:IUnit 1, Reactor Vessel - Filling and Venting the RCS38
OP A-2:IXReactor Vessel - Vacuum Refill of the RCS6
OP D-1:VAuxiliary Feedwater System - Alternate AuxiliaryFeedwater Supplies
ProceduresNumberTittleRevisionAttachmentA-9OP L-1Plant Heatup From Cold Shutdown to Hot Standby73OP O-25Operability Order O-25 Advisory on Equipment AvailabilityAfter a Major Earthquake

3OP1 .ID2Time Critical Operator Actions1AECG 23.6480VAC Class 1 Switchgear Ventilation System1

ECG 23.1Area Temperature Monitoring2
MP I-2.5-1Calibration of Plant Pressure Gauges14
STP I-1ARoutine Shift Checks Required by Licenses101
STP I-1ARoutine Shift Checks Required by Licenses101B
MP I-2.5-1Calibration of Plant Pressure Gauges22
OM7.ID1Problem Identification and Resolution - Action Requests22
CF6.NE1Instrument Channel Uncertainty and SetpointMethodologies
2AMA1.DC11Risk Assessment 7AWP E-022Check Valve Data Evaluation and Trending Procedure withthe Viper System

0MA1 .ID6Check Valve Maintenance, Testing, and InspectionProgram 2OP

J-6BDiesel Generators8MP M-4.27Main Steam Isolation Valve Disassembly, Inspection and Reassembly
13STP V-18Check Valve Inspection18STP V-7BTest of Engineered Safeguards, Valve Interlocks and RHRPump Trip from RWST Level Channels
23AR PK02-16RHR System23OP
AP-16Malfunction of the RHR System13
OP
AP-24Shutdown LOCA6
OP J-2:IIIStartup Bank - Shutdown and Clearing20
EOP
ECA-0.3Restore 4 kV Buses13
OP A-4A:1Pressurizer - Make Available20
ProceduresNumberTittleRevisionAttachmentA-10DCM No. T-21Grounding4ADCM No. S-634160 System13ADCM No. T-18Electrical System Protection10BRecurring Task ActivityR0270371-01R0270369-01R0262758-01R0263108-01R0263109-01R0282308-01R0258244-01R0225115-01R0236401-01R0285241-01
R0285243-01R0236400-01R0250003-01R0237109-01R0285242-01
Work OrdersR0181183R0220536R0192663R0192661R0192559R0201243R0201104R0232469C0203299C0202623
A0455919A0532178A0478550A0478546C0192576
A0478324A0494967Q0003928C0156366C0178210
C0196490R0218005R0227043R0252440R0292390
R0217862R0218006R0242100R0255373R0292787Miscellaneous DocumentsNumberTittleRevision/DateDCM S-3BDesign Criteria
Memorandum Auxiliary FeedwaterSystem 15ADCM S-14Design Criteria
Memorandum Component Cooling WaterSystem 15EDCM S-17BDesign Criteria
Memorandum Auxiliary Salt WaterSystem 18IEN 04-01Auxiliary Feedwater Pump Recirculation Line OrificeFouling - Potential Common Cause Failure at Point
BeachJanuary 2004CRANE Report2-CCW-601 Non-Intrusive Testing Analysis July 16, 2002NP-6815-DDetection and Control of Microbiologically InducedCorrosion 1990

Miscellaneous

DocumentsNumberTittleRevision/DateAttachmentA-11ABCA-00597Sample Analysis ReportNovember16, 2004ABCA-00628Sample Analysis ReportJanuary 26, 2005ABCA-00681aSample Analysis ReportNovember17, 2005ABCA-00681Sample Analysis ReportNovember17, 20052005-S014-009Remove CCW Supply/return Piping from AbandonedWaste Concentrator Package Coolers (Unit 0)

(Long Term Plan)ICEA Pub. S-66-

24NEMA Standard for Cross-Linked ThermosettingPolyethylene Insulated Wire & Cable for the Transmission & Distribution of Electrical Energy, NEMA
Pub. No. WC 7-1982
4NEMA Std.. No.WC 8-1992Ethylene Propylene Rubber Insulated Wire & Cable 3Okonite ProductDataOkoguard Okolon Type
MV-105, 5/8 kV Shielded Power Cable9/29/06Altran Report02810-TR-001Component Cooling Pump 4 kV Cable Evaluation1TS3.ID2, Att. 8.1Supplemental LBIE Screen per AR A0677849ASCO CatalogNo.
NP-13 and 4 Way Solenoid Valves
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