IR 05000361/2011010

IR 05000361-11-010, 05000362-11-010, on 06/20/2011 - 09/13/2011, San Onofre Nuclear Generating Station, Units 2 and 3, Baseline Inspection, NRC Inspection Procedure 71111.21, Component Design Basis Inspection.

ML112870563
Person / Time
Site:	San Onofre
Issue date:	10/14/2011
From:	Thomas Farnholtz Region 4 Engineering Branch 1
To:	Peter Dietrich Southern California Edison Co
References
IR-11-010
Download: ML112870563 (59)

Text

October 14, 2011

Mr. Peter Dietrich Senior Vice President and Chief Nuclear Officer Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92674-0128

SUBJECT: SAN ONOFRE NUCLEAR GENERATING STATION - NRC COMPONENT DESIGN BASES INSPECTION NRC REPORT 05000361/2011010 and 05000362/2011010

Dear Mr. Dietrich:

On September 13, 2011, the US Nuclear Regulatory Commission (NRC) completed a Component Design Bases Inspection at your San Onofre Nuclear Generating Station, Units 2 and 3 facility. The enclosed report documents our inspection findings. The preliminary findings were discussed on July 22, 2011, with Mr. P. Dietrich, Senior Vice President & Chief Nuclear Officer and other members of your staff. After additional in-office inspection, a final telephonic exit meeting was conducted on September 13, 2011, with Mr. R. St. Onge, Director, Nuclear Regulatory Affairs, and others of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license. The team reviewed selected procedures and records, observed activities, and interviewed cognizant plant personnel.

Based on the results of this inspection, the NRC has identified six findings that were evaluated under the risk significance determination process. Violations were associated with all of the findings. All six of the findings were found to have very low safety significance (Green) and the violations associated with these findings are being treated as noncited violations, consistent with the NRC Enforcement Policy.

If you contest any of the noncited violations, or the significance of the violations you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the US Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 612 East Lamar Blvd., Suite 400, Arlington, Texas 76011; Southern California Edison Company - 2 -

the Director, Office of Enforcement, US Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the San Onofre Nuclear Generating Station, Units 2 and 3 facility. The information you provide will be considered in accordance with Inspection Manual Chapter 0305. In addition, if you disagree with the characterization of the cross-cutting aspect assigned to any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region IV, and the NRC Resident Inspector at San Onofre Nuclear Generating Station, Units 2 and 3 facility.

In accordance with Code of Federal Regulations, Title 10, Part 2.390 of the NRC's Rules of Practice, a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document 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/

Thomas R. Farnholtz, Chief Engineering Branch 1 Division of Reactor Safety

Docket Nos. 50-361, 50-362 License Nos. NPF-10, NPF-15

Enclosure:

NRC Inspection Report 05000361/2011010 and 05000362/2011010

w/Attachment:

1 - Supplemental Information cc w/

Enclosure:

Distribution via Listserv for SONGS

Southern California Edison Company - 3 -

Electronic distribution by RIV: Regional Administrator (Elmo.Collins@nrc.gov) Deputy Regional Administrator (Art.Howell@nrc.gov) DRP Director (Kriss.Kennedy@nrc.gov) DRS Director (Anton.Vegel@nrc.gov) DRS Deputy Director (Tom.Blount@nrc.gov) Senior Resident Inspector (Greg.Warnick@nrc.gov) Resident Inspector (John.Reynoso@nrc.gov) Branch Chief, DRP/D (Ryan.Lantz@nrc.gov) Senior Project Engineer, DRP/D (Don.Allen@nrc.gov) SONGS Administrative Assistant (Heather.Hutchinson@nrc.gov) Project Engineer, DRP/D (David.You@nrc.gov) Project Engineer, DRP/D (Brian.Parks@nrc.gov) Public Affairs Officer (Victor.Dricks@nrc.gov) Public Affairs Officer (Lara.Uselding@nrc.gov) Project Manager (Randy.Hall@nrc.gov) Acting Branch Chief, DRS/TSB (Dale.Powers@nrc.gov) RITS Coordinator (Marisa.Herrera@nrc.gov) Regional Counsel (Karla.Fuller@nrc.gov) Congressional Affairs Officer (Jenny.Weil@nrc.gov) OEMail Resource

Inspection Reports/MidCycle and EOC Letters to the following: ROPreports Only inspection reports to the following:

RIV/ETA: OEDO (Mark.Franke@nrc.gov) DRS/TSB STA (Dale.Powers@nrc.gov)

R:\REACTORS\SONGS\SO 2011010 CDBI-RAK.docx ML ADAMS No x Yes x SUNSI Review CompleteReviewer Initials: TRFCategory B.1 Publicly Availablex Non-sensitiveCategory A Non-publicly Available SensitiveKEYWORD: A.X Sensitive

-XXXR4:DRS EB1 PSB2:SRI EB2:RIR1: EB1:SRI C: EB1 R. Kopriva J. Drake J. WatkinsS. PindaleT. Farnholtz/RA/ /RA/ RAK for /RA/ RAK for /RA/ RAK for /RA/ 10/13/2011 10/13/2011 10/13/2011 10/13/2011 10/14/2011 OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax Enclosure ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION REGION IV Docket: 50-361, 50-362 License: NPF-10, NPF-15 Report: 05000361/2011010 and 05000362/2011010 Licensee: Southern California Edison Co. (SCE) Facility: San Onofre Nuclear Generating Station, Units 2 and 3 Location:

5000 S. Pacific Coast HwySan Clemente, California Dates: June 20, 2011 through September 13, 2011Team Leader:

R. Kopriva, Senior Reactor Inspector, Engineering Branch 1, Region IV Inspectors:

J. Drake, Senior Reactor Inspector, Plant Support Branch 2, Region IVJ. Watkins, Reactor Inspector, Engineering Branch 2, Region IV S. Pindale, Senior Reactor Inspector, Engineering Branch 1, Region I Accompanying Personnel:

H. Campbell, Ph.D., Mechanical Contractor, Beckman and AssociatesS. Kobylarz, Electrical Contractor, Beckman and Associates W. Sherbin, Mechanical Contractor, Beckman and Associates Approved By:

Thomas R. Farnholtz, Branch ChiefEngineering Branch 1 Enclosure

SUMMARY OF FINDINGS

IR 05000361/2011010, 05000362/2011010; 06/20/2011 - 09/13/2011; San Onofre Nuclear Generating Station, Units 2 and 3, baseline inspection, NRC Inspection Procedure 71111.21, "Component Design Basis Inspection."

The report covers an announced inspection by a team of four regional inspectors and three contractors. Six findings were identified. All of the findings were of very low safety significance. The final significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 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 4, dated December 2006.

Cornerstone: Mitigating Systems

Green.

The team identified a Green noncited violation of 10 CFR 50, Appendix B, Criterion III, "Design Control," which states in part: "Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions." Specifically, from initial construction until July 22, 2011, the licensee had not properly evaluated the seismic qualification of the buried emergency diesel generator fuel oil storage tanks, to ensure that the tank's structures would not fail during a seismic event. The calculation did not accurately reflect the actual installed condition of the fuel oil tanks. The team determined that failure of the tanks to remain intact would impact the capability of the safety related emergency diesel generators to perform their design function following the event. This finding was entered into the licensee's corrective action program as Nuclear Notification NN-201548802.

The team determined that the failure to have an adequate seismic calculation for emergency diesel generator fuel oil storage tanks was a performance deficiency. The finding was more than minor because it was associated with the mitigating systems cornerstone attribute of design control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inadequate design analysis of these components could have resulted in structural failure, preventing continued operation of the emergency diesel generators after an earthquake. In accordance with Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the team determined that the finding was of very low safety significance (Green) because it did not represent a design issue resulting in the loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. Specifically, the licensee performed a preliminary re-evaluation of the tank shell stresses due to the concrete structures, and determined that the tank stresses were still within the American Society of Mechanical Engineers (ASME) Code allowable stresses following a Safe Shutdown Earthquake (SSE). The team reviewed the evaluation, and concurred that the stresses were below those allowed by ASME Boiler and Pressure Vessel Code. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance (Section 1R21.2.9).

Green.

The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," which states in part: "Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions." Specifically, as of July 22, 2011, the licensee failed to evaluate that the voltage available at the emergency diesel generator diesel engine air start solenoid would be sufficient to ensure starting the diesel engine under design basis conditions. The licensee failed to incorporate the required minimum operating voltage for the emergency diesel generator air start solenoids into Calculation E4C-017. This finding was entered into the licensee's corrective action program as Nuclear Notifications NN-201513266 and NN-201566686.

The team determined that the failure to incorporate the required minimum operating voltage for the emergency diesel generator air start solenoids into Calculation E4C-017 was a performance deficiency. The finding was more than minor because it was associated with the mitigating systems cornerstone attribute of design control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the team determined that the finding was of very low safety significance (Green) because it did not represent a design issue resulting in the loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. Specifically, the licensee performed subsequent analyses and actual tests of the air start solenoids, which demonstrated that the emergency diesel generator air start solenoids would function as required to mitigate an accident. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance (Section 1R21.2.13).

Green.

The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, "Design Control," which states in part: "Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions." Specifically, as of July 22, 2011, the licensee did not incorporate the vendor required amperage limit, identified in engineering change package ECP-040201281-2, for bus load limit requirements, into 480 Volt Switchgear Operating Instructions SO23-6-3. This finding was entered into the licensee's corrective action program as Nuclear Notification NN201570846.

The team determined that the failure to incorporate the vendor required amperage limit, identified in engineering change package ECP-040201281-2, for bus load limit requirements, into 480 Volt Switchgear Operating Instructions SO23-6-3 was a performance deficiency. The finding was more that minor because it was associated with the mitigating systems cornerstone attribute of design control, to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the finding was determined to have very low safety significance (Green) because it did not represent a design issue resulting in the loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. Specifically, the licensee had never implemented 480 Volt Switchgear Operating Instructions SO23-6-3 for the purpose of cross tying busses in an emergency, where the limiting load on the bus may have been exceeded. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance (Section 1R21.2.14).

Green.

The team identified a Green noncited violation of 10 CFR 50, Appendix B, Criterion III, "Design Control," which states in part: "Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions." Specifically, as of July 22, 2011, the licensee failed to incorporate the fuse resistance, fuse clips resistance, and cable temperature and resistance effects (for Auxiliary Feedwater High Energy Line Breaks only), into Calculations E4C-084 and E4C-085, for degraded voltage conditions. This finding was entered into the licensee's corrective action program as Nuclear Notification NN-201546570 and NN-201550186.

The team determined that the failure to fully evaluate the circuit load in determining design limits in electrical calculations for degraded voltage conditions was a performance deficiency. The finding was more than minor because it was associated with the mitigating systems cornerstone attribute of design control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the team determined that the finding was of very low safety significance (Green) because it did not represent a design issue resulting in the loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. Specifically, the licensee performed subsequent preliminary analyses which demonstrated that the control circuits, where marginal voltage was available, would function as required to mitigate an accident. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance (Section 1R21.2.15).

Green.

The team identified a Green noncited violation of 10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," which states in part: "Each holder of a license to operate a nuclear power plant shall monitor the performance or condition of structures, systems, or components, against licensee-established goals, in a manner sufficient to provide reasonable assurance that such structures, systems, and components, are capable of fulfilling their intended functions, and when the performance or condition of a system, structure, or component, does not meet established goals, appropriate corrective actions shall be taken." Specifically, as of July 22, 2011, the licensee failed to adequately monitor the condition of the Flood Level Detecting system in a manner to provide reasonable assurance the system could perform its intended function. The licensee failed to properly evaluate Maintenance Rule Functional Failures and take appropriate corrective actions to improve system performance. These level switches are connected to control room annunciation to warn the control room of flooding in a space that has safety-related or important to safety components. This has been entered into the licensee's corrective action program as Nuclear Notifications NN-201567315 and NN-201570575.

The team determined that the failure to properly maintain the flood level sensors which are used for control room annunciation to warn the control room of flooding of a space that has safety related or important to safety components, was a performance deficiency. The finding was more than minor because it was associated with the mitigating systems cornerstone attribute of design control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not maintain flood level sensors appropriately to provide reasonable assurance that the components would be capable of fulfilling their intended function. In accordance with Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the team determined that the finding represented the degradation of equipment and functions specifically designed to provide notification to the control room of flooding of spaces with safety related or important to safety equipment and components. Therefore, the finding was potentially risk significant and a Phase 3 analysis was required. The preliminary significance determination was based on Inspection Manual Chapter 0609, Appendix M, "Significance Determination Process Using Qualitative Criteria." The senior reactor analyst determined qualitatively that the risk was very low for the following reasons: (1) the frequency of internal flooding is very low, (2) floods in most of the problem areas would not cause a transient, (3) redundant indications of flooding exist, including fire and sump pump operations, and (4) none of the potentially flooded areas would likely affect more than one train of safety equipment. This finding involved a cross-cutting aspect in the area of Human Performance, Resources, because the licensee failed to assure that equipment and other resources were available and adequate to assure nuclear safety. Specifically, the licensee was not able to maintain the flood level switches adequately to assure nuclear safety due to long-standing equipment issues H.2(a)(Section 1R21.3.2).

Green.

The team identified a Green non-cited violation, with multiple examples, of 10 CFR 50, Appendix B, Criterion VI, "Document Control," which states in part: "Measures shall be established to control the issuance of documents, such as instructions, procedures, and drawings, including changes thereto, which prescribe all activities affecting quality. These measures shall assure that documents, including changes, are reviewed for adequacy and approved for release." Specifically, on June 23, 2011, the team identified numerous drawing inconsistencies where changes to certain components were not changed on all affected drawings and procedural errors where changes were not made to all affected documents. The licensee has entered the errors into their corrective action program under numerous Nuclear Notifications listed in section 4AO2.

The team identified that collectively, from a program perspective, the failure to properly incorporate design changes of components in the plant to all affected drawings, procedures, or instructions, was a performance deficiency. The finding was more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. In accordance with Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the team determined that the finding was of very low safety significance (Green) because it did not represent a design issue resulting in the loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. Specifically, none of the documents with the identified errors had been used in response to any events or plant perturbations. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance (Section 4OA2).

B. Licensee-Identified Violations

No finding were identified.

REPORT DETAILS

REACTOR SAFETY

Inspection of component design bases verifies the initial design and subsequent modifications 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 important design features may be altered or disabled during modifications. 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

To assess the ability of the San Onofre Nuclear Generating Station equipment and operators to perform their required safety functions, the team inspected risk significant components, and the licensee's responses to industry operating experience. The team selected risk significant components for review, using information contained in the San Onofre Nuclear Generating Station Probabilistic Risk Assessment and the U. S. Nuclear Regulatory Commission's (NRC) standardized plant analysis risk model. In general, the selection process focused on components that had a risk achievement worth factor greater than 1.3 or a risk reduction worth factor greater than 1.005. The items selected included components in both safety-related and nonsafety related systems including pumps, circuit breakers, heat exchangers, transformers, and valves. The team selected the risk significant operating experience to be inspected based on its collective past experience.

.1 Inspection Scope

To verify that the selected components would function as required, the team reviewed design basis assumptions, calculations, and procedures. In some instances, the team performed calculations to independently verify the licensee's 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, and industry operating experience records 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, as well as during simulated actions in the plant.

The team performed a margin assessment and detailed review of the selected risk-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 modifications, and 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; 0 CFR 50.65(a)1 status; operable, but degraded, conditions; NRC resident inspector input of problem equipment; system health reports; industry operating experience; 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 to 25 samples that include risk-significant and low design margin components, containment related components, and operating experience issues. The sample selection for this inspection was 18 components, one of which is containment related, two operating experience items, and two Event Scenario-Based activities. The selected inspection and associated operating experience items supported risk significant functions including the following:

a. Electrical power to mitigation systems: The team selected several components in the offsite and onsite electrical power distribution systems to verify operability to supply alternating current (AC) and direct current (DC) power to risk significant and safety-related loads in support of safety system operation in response to initiating events such as loss of offsite power, station blackout, and a loss-of-coolant accident with offsite power available. As such the team selected:

• Emergency Diesel Generator Emergency Supply Fan S21503MA274

• 4160 Volt Bus 2A06 to Bus 3A06 Cross-Tie

• Emergency Diesel Generator 2G002 Start and Trip Functions

• 480 Volt Load Center Transformer 3B06X b.

Seismic concern on components: The team reviewed several components required to minimize the effects of seismic activity as an initiating event. These components were required to provide cooling and mitigate the consequences of analyzed events. As such the team selected:

• Emergency Diesel Generator Fuel Oil Tank - TO 35 (S22421MTO35)

• Salt Water Outfall (Discharge)

• Component Cooling Water Surge Tank (S21203MT004)c. Mitigating systems needed to attain safe shutdown. The team reviewed components required to perform the safe shutdown of the plant. As such the team selected:

• Emergency Core Cooling System Suction Header Check Valve S21204MU001

• Reactor Coolant Pump P001 Seal Heat Exchanger

• Refueling Water Tank Outlet Valve 2HV9301

.2.1 Unit 3 Low Pressure Safety Injection Pump (P016)

a. Inspection Scope

The team reviewed portions of the Updated Final Safety Analysis Report, Technical Specifications, system description and design bases documents to determine the system design and performance criteria for the Unit 3 Low Pressure Safety Injection pump P016. The team also performed a walkdown of the pump area to examine the installed configuration and general material condition of the pump. Further the team held discussions with cognizant licensee individuals with focus on pump performance and testing procedures. Specifically the team reviewed:

• Piping and instrument diagrams of the as built pump configuration, and associated flow, pressure and temperature instruments

• Specifications of the orifice plates used to determine Low Pressure Safety Injection flow during performance of Inservice Test (IST) surveillances

• Instrument uncertainty Calculations which evaluated Instruments used in the Inservice Testing Program, (shutdown cooling temperature, suction and discharge pressure indicators, flow indications)

• Calculations used to evaluate Inservice Testing and design basis test performance requirements, (includes instrument uncertainties coupled with performance requirements)

• Quarterly and Full Flow surveillances procedures and test results used to verify required Residual Heat Removal pump performance and potential pump degradation

• Nuclear Notifications (corrective action documents) addressing pump performance, maintenance and operability evaluations

b. Findings

No finds were identified.

.2.2 Unit 2 Salt Water Cooling Pump

a. Inspection Scope

The team reviewed the Updated Final Safety Analysis Report, Technical Specifications, system design criteria to determine the system design and performance criteria for the Unit 2 Salt Water Cooling Pump P307. Further, selected drawings, operating procedures, and past Nuclear Notification reports for pump P307 were reviewed. After a walkdown of the Salt Water Cooling pump and nearby areas, the team discussed the current health and condition of the pump with the system engineer. Specifically the team reviewed:

• Piping and instrument diagrams of the as built pump configuration and associated flow, pressure and temperature instruments

• Schematics of the intake structure, including the layout of the Salt Water Cooling and Circulating Water pumps

• Instrument uncertainty calculations, which evaluated Instruments, used in the Inservice Testing Program, (shutdown cooling temperature, suction and discharge pressure indicators, flow indications)

• Quarterly and Full Flow surveillances procedures and test results used to verify required Residual Heat Removal pump performance and potential pump degradation

• Nuclear Notifications addressing pump performance, maintenance and operability evaluations

b. Findings

No findings were identified.

.2.3 Reactor Coolant Pump P001 Seal Heat Exchanger

a. Inspection Scope

The team reviewed the Updated Final Safety Analysis Report, Reactor Coolant System Design Basis Document and Reactor Coolant System Description, and selected drawings for the Reactor Coolant Pump P001 Seal Heat Exchanger to determine the design and performance criteria for the Reactor Coolant Pump P001 Seal Heat Exchanger. Further, the team discussed the history of the major change in design and implementation of the current seal heat exchangers with the cognizant system engineers. Specifically, the team reviewed:

• System Health reports of the reactor coolant system with focus on the reactor coolant pumps

• Nuclear Notifications leading to decisions for implementing design changes to the reactor coolant pump seal heat exchangers

• Piping and instrument diagrams of the reactor coolant pumps and associated seal heat exchangers

• Sectional drawings of the currently installed as-built reactor coolant pump seal heat exchangers

• Design change package, which implemented removal of the high temperature component cooling water return lockout circuitry

b. Findings

No findings were identified.

.2.4 Refueling Water Tank Outlet Valve (2HV9301)

a. Inspection Scope

The team reviewed the Updated Final Safety Analysis Report, system design criteria, selected drawings, and operating and test procedures for refueling water tank outlet motor-operated valve (MOV) 2HV9301. The team also performed walkdowns, held discussions with cognizant licensee individuals, and reviewed corrective action documents. Specifically, the team reviewed:

• Motor-operated valve diagnostic and stroke-time test results

• Calculations that determined motor-operated valve settings such as torque and limit switch settings

• The motor-operated valve weak link calculations to ensure the ability of the motor-operated valves to remain structurally functional while stroking under design basis conditions

• Valve design and operating analyses to determine maximum differential pressure expected across the valves during worst case operating conditions

• Thermal binding and pressure locking analyses

• Degraded voltage conditions, thermal overload sizing, and voltage drop calculation results to confirm that the motor-operated valves would have sufficient voltage and power available to perform their safety function at degraded voltage conditions

• Vendor manuals to ensure adequate maintenance and operation of the valves

• Inservice Test basis documents and associated test results

• Maintenance activities to ensure the components were being maintained in accordance with vendor recommendations

b. Findings

No findings were identified.

.2.5 Emergency Core Cooling System Suction Header Check Valve (S21204MU001)

a. Inspection Scope

The team reviewed the Updated Final Safety Analysis Report, system design criteria, selected drawings, and maintenance and test procedures for emergency core cooling system suction header check valve S21204MU001. The team also performed walkdowns, held discussions with cognizant licensee individuals, and reviewed corrective action documents. Specifically, the team reviewed:

• Inservice Test basis documents and associated test results, including forward flow and leakage rate test results

• Calculations that provided the bases for the IST test criteria

• Check valve internal inspection/examination results

• Vendor manuals and associated maintenance activities to ensure the valves were being maintained in accordance with vendor recommendations

b. Findings

No findings were identified.

.2.6 Auxiliary Feedwater Check Valve S21305MU124 (for SG089) from Motor-Driven

Pump 141

a. Inspection Scope

The team reviewed the Updated Final Safety Analysis Report, system design criteria, selected drawings, and maintenance and test procedures for auxiliary feedwater check valve S21305MU124. The team also held discussions with cognizant licensee individuals and reviewed corrective action documents. Specifically, the team reviewed:

• Inservice test basis documents and associated test results, including forward flow and leakage rate test results

• Calculations that provided the bases for the inservice test criteria

• Check valve internal inspection/examination results

• Vendor manuals and associated maintenance activities to ensure the valves were being maintained in accordance with vendor recommendations

b. Findings

No findings were identified.

.2.7 Containment Emergency Sump Outlet Isolation Valve (2HV9302)

a. Inspection Scope

The team reviewed the Updated Final Safety Analysis Report, system design criteria, selected drawings, and operating and test procedures for containment emergency sump outlet motor-operated valve 2HV9302. The team also performed walkdowns, held discussions with cognizant licensee individuals, and reviewed corrective action documents. Specifically, the team reviewed:

• Motor-operated valve diagnostic, logic and stroke-timing test results

• Calculations that determined motor-operated valve settings such as torque and limit switch settings

• The motor-operated valve weak link calculations to ensure the ability of the motor operated valves to remain structurally functional while stroking under design basis conditions

• Valve design and operating analyses to determine maximum differential pressure expected across the valves during worst case operating conditions

• Degraded voltage conditions, thermal overload sizing, and voltage drop calculation results to confirm that the motor-operated valves would have sufficient voltage and power available to perform their safety function at degraded voltage conditions

• Vendor manuals to ensure adequate maintenance and operation of the valves

• Inservice test basis documents and associated test results

• Maintenance activities to ensure the components were being maintained in accordance with vendor recommendations

b. Findings

. No findings were identified.

.2.8 Emergency Diesel Generator Emergency Supply Fan (S21503MA274)

a. Inspection Scope

The emergency diesel generator building supply fan operates to provide ambient air to remove heat generated by the emergency diesel generator and auxiliaries. The team reviewed the system design criteria, selected drawings, test and maintenance requirements for the emergency diesel generator building supply fan. The team also performed walkdowns, and held discussions with cognizant licensee individuals. Specifically the team reviewed:

• Piping and instrumentation drawings

• Fan sizing calculation to ensure adequate air flow for heat removal

• Fan start/stop control logic

• Seismic qualification of the fan, and associated ductwork

• Required specifications for fan and ductwork

• Completed surveillances of Emergency Diesel Generator to ensure fan operation was satisfactory

• Airflow measurement test to ensure fan capacity was in accordance with design airflow

b. Findings

No findings were identified.

.2.9 Emergency Diesel Generator Fuel Oil Tank (S22421MTO35)

a. Inspection Scope

The emergency diesel generator fuel oil storage tanks are buried structures, sized to provide seven days of fuel oil for the site emergency diesel generators. There are four tanks for both units, with one tank per emergency diesel generator. The team reviewed the system design criteria, selected drawings, and maintenance requirements for the emergency diesel generator fuel oil storage tank. The team also performed walkdowns, and held discussions with cognizant licensee individuals. Specifically the team reviewed:

• Piping, instrumentation and structural drawings

• Fuel oil tank sizing related to meeting Technical Specification requirements for adequate volume

• Instrumentation calculations and completed calibrations for tank level measurement

• Seismic qualification of the tank

• Required specifications for buried tank

• Completed surveillances performed to satisfy Technical Specifications Surveillance Requirements for oil sampling, sediment, and tank internal inspection

• Required fuel oil transfer pump submergence to prevent vortexing

• Seismic interaction between the buried tank and adjacent structures

b. Findings

Inadequate Assessment of Seismic Qualification of Emergency Diesel Generator Buried Fuel Oil Tanks

Introduction.

The team identified a Green noncited violation of 10 CFR 50, Appendix B, Criterion III, "Design Control," because the licensee had not properly evaluated the seismic qualification of the buried emergency diesel generator fuel oil storage tanks, to ensure that the tank's structures would not fail during a seismic event. The calculation did not accurately reflect the actual installed condition of the fuel oil tanks.

Description.

The four emergency diesel generator fuel storage tanks are located underground. Each has a capacity of 55,000 gallons. During a walkdown of the Unit 2 and Unit 3 buried fuel oil tank area, the team observed concrete structures above the tanks. Above one end of each tank there is a concrete vault, with the roof above plant grade, which houses the instruments, fuel oil transfer pumps, and one of the manways. Above the other end of the tank is a concrete structure housing the filling station, and another manway. Following the walkdown, the team requested the licensee to provide the seismic analysis for the tanks, including the seismic interaction analysis of the concrete structures installed above the tanks. The team was given the tank vendor seismic evaluation, calculation number SO23-407-7-9, "Seismic Design Analysis of Diesel Fuel Oil Storage Tanks." It states that the tanks are designed to Seismic Class I for nuclear service in accordance with American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section III, Subsection ND, for Class 3 components. The team reviewed the evaluation, and determined that the evaluation was non-conservative for the following reasons:

• The calculation stated that the tanks were surrounded 100 percent by soil, and that the load was uniformly distributed. Actually, there are two concrete structures of substantial weight on top of each end of the tanks, with about 9 inches of soil between the tank top and the structures.

• The stress analysis of the tank shell assumed the tank centerline was buried to a depth of 11 feet of soil. The centerline is actually buried under 16 feet of soil. This would increase the soil pressure by about 50 percent over what was evaluated.

The team was concerned that the seismic evaluation of record did not represent the earthquake loading conditions of the installed tanks. As a result of the team's concern, the licensee entered the issue into their corrective action program as Nuclear Notification NN-201548802, and performed a preliminary re-evaluation of the tank shell stresses due to the concrete structures, and determined that the tank stresses are still within the ASME Code allowable stresses following a Safe Shutdown Earthquake. The team reviewed the evaluation, and concurred that the stresses are below those allowed by ASME Code. A re-evaluation of the stresses due to the additional five feet of soil loading was not necessary because the calculation stated that there is 100 percent margin in the tank stresses due to soil loading, and the actual soil loading increase was approximately 50 percent.

Analysis.

The team determined that the failure to have an adequate seismic calculation for the emergency diesel generator fuel oil storage tanks was a performance deficiency. The finding was more than minor because it was associated with the mitigating systems cornerstone attribute of design control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inadequate design analysis of these components could have resulted in structural failure, preventing continued operation of the emergency diesel generators after an earthquake. In accordance with Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the team determined that the finding was of very low safety significance (Green) because it did not represent a design issue resulting in the loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. Specifically, the licensee performed a preliminary re-evaluation of the tank shell stresses due to the concrete structures, and determined that the tank stresses were still within the ASME Code allowable stresses following a Safe Shutdown Earthquake. The team reviewed the evaluation, and concurred that the stresses were below those allowed by ASME Boiler and Pressure Vessel Code. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.

Enforcement.

The team identified a Green noncited violation of 10 CFR 50, Appendix B, Criterion III, "Design Control," which states in part: "Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions." Contrary to the above, the licensee failed to assure that regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, from initial construction until July 22, 2011, the licensee had not properly evaluated the seismic qualification of the buried emergency diesel generator fuel oil storage tanks, to ensure that the tank's structures would not fail during a seismic event. The calculation did not accurately reflect the actual installed condition of the fuel oil tanks. The team determined that failure of the tanks to remain intact would impact the capability of the safety related emergency diesel generators to perform their design function following the event. This finding was entered into the licensee's corrective action program as Nuclear Notification NN-201548802. Because this violation was of very low significance (Green) and has been entered into the licensee's corrective action program, this violation is being treated as a noncited violation consistent with the NRC Enforcement Policy: NCV 05000361/2011010-01, "Inadequate Assessment of Seismic Qualification of Emergency Diesel Generator Buried Fuel Oil Tanks."

.2.10 Component Cooling Water Surge Tank

(S21203MT004)

a. Inspection Scope

The component cooling water surge tank provides an inventory of water at sufficient pressure to provide the component cooling water pump net positive suction head (NPSH), and sufficient water for component cooling water system leakage makeup. The team reviewed the system design criteria, selected drawings, test and maintenance requirements for the component cooling water surge tank. The team also performed walkdowns, and held discussions with cognizant licensee individuals. Specifically the team reviewed:

• Piping and instrumentation drawings

• Surge tank sizing related to meeting component cooling water leakage requirements, and meeting component cooling water pump net positive suction head requirements

• Instrumentation calculations for tank pressure measurement instruments

• Seismic qualification of the tank

• Required specifications for component cooling water tank

• Completed surveillances performed to satisfy Technical Specifications Surveillance Requirements for backup nitrogen supply capability

b. Findings

No findings were identified.

.2.11 Salt Water Outfall (Discharge)

a. Inspection Scope

The saltwater outfall is the discharge line to the ultimate heat sink (Pacific Ocean) for the saltwater cooling system. An emergency discharge line, common to Units 2 and 3, is provided in the event of blockage of the normal discharge line. The team reviewed the system design criteria, selected drawings, test and maintenance requirements for the Salt Water Outfall. The team also performed walkdowns of the outfall area, and held discussions with cognizant licensee individuals. Specifically the team reviewed:

• Piping and structural drawings

• Tsunami evaluation of the outfall area

• Evaluation that determined seismic failure of the outfall would not block safety-related salt water cooling flow

• Operating and maintenance procedures for the emergency discharge line, and valves

b. Findings

No findings were identified.

.2.12 4160 Volt Bus 2A06 to Bus 3A06 Cross-Tie

a. Inspection Scope

The team reviewed the system design criteria, selected drawings, design calculations and operating procedures for the Bus 2A06 to Bus 3A06 cross-tie. The team also performed walkdowns, reviewed Nuclear Notifications (condition reports) for recurrence of adverse conditions affecting reliability, and held discussions with cognizant licensee individuals. Specifically, the team reviewed:

• One line diagrams of the cross-tie breakers and cable

• Schematic diagrams for the cross-tie breakers

• Surveillance and maintenance tests for the cross-tie

• Operations test for the cross-tie

• Cross-tie cable sizing calculation

• Design basis load requirement for cross-tie

• Cross-tie breaker protective relay setting calculation

b. Findings

No findings were identified.

.2.13 Emergency Diesel Generator 2G002 Start and Trip Functions

a. Inspection Scope

The team reviewed selected emergency diesel generator operating devices that under design basis conditions provided for starting the diesel engine and also that provided for the generator breaker to close and automatically trip. The team also performed walkdowns, reviewed Nuclear Notifications for recurrence of adverse conditions affecting reliability, and held discussions with cognizant licensee individuals. Specifically, the team reviewed:

• Updated Final Safety Analysis Report for emergency diesel generator start and trip control devices

• Engine Systems, Inc. Report No. 10CFR21-0077, Graham-White Air Start Solenoid Valves #712-015 & #712-065

• System one line and schematic diagrams

• Vendor rating specifications for diesel engine air start solenoids and generator breaker close and trip devices

• Calculated voltage available at diesel generator air start solenoids and generator breaker close and trip devices

• Substitution Equivalency Evaluation (SEE) for Unit 2/3 for Emergency Diesel Generator Air Start Solenoid Valves

• Surveillance test results of diesel generator air start solenoids and generator breaker close device and trip devices enabled during accident conditions

b. Findings

Failure to evaluate that sufficient voltage would be available at the Emergency Diesel Generator air start solenoid

Introduction.

The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," because the licensee failed to evaluate that the voltage available at the emergency diesel generator diesel engine air start solenoid would be sufficient to ensure starting the diesel engine under design basis conditions.

Description.

The team performed walkdowns of Emergency Diesel Generator 2G002 to verify the vendor nameplate data for the generator and the diesel engine air start solenoid valves. The diesel engine air start solenoid valves were evaluated by the licensee under SEE No. 990010, based on Engine Systems, Inc. Report No. 10CFR21-0077 that resulted from NRC 10 CFR Part 21 Report 1998-12-0. The team also reviewed design calculation E4C-017.1, Class 1E 125 Vdc System Data/Loading, to verify the adequacy of the voltage available at the engine air start solenoid valve terminals during design basis conditions. The licensee had calculated transient voltage available (between approximately 81 and 85 Vdc) during start attempts of the diesel engine. The team found that the calculated available voltage to the solenoid was less than the minimum value specified on the solenoid valve vendor nameplate, 90 Vdc, and that no analysis had been performed within the licensee's calculation to justify the adequacy of the available voltage. The team found that the licensee failed to include vendor cut sheets in the calculation that indicated the minimum operating voltage and determined that the licensee did not evaluate the availability of sufficient voltage at the solenoids during design basis conditions. During the inspection, the licensee tested spare solenoid valves from their warehouse that were the same type and model as the solenoid valves that were installed on the emergency diesel generators. The testing results were satisfactory, providing reasonable assurance of operability for the solenoid valves during transient conditions when there would be less than rated voltage available. In addition, the licensee performed a preliminary calculation that determined that the minimum continuous voltage available at the solenoid would meet the manufacturer's tested and specified minimum value.

Analysis.

The team determined that the failure to incorporate the required minimum operating voltage for the emergency diesel generator air start solenoids into Calculation E4C-017 was a performance deficiency. The finding was more than minor because it was associated with the mitigating systems cornerstone attribute of design control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the team determined that the finding was of very low safety significance (Green) because it did not represent a design issue resulting in the loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. Specifically, the licensee performed subsequent analyses and actual tests of the air start solenoids, which demonstrated that the emergency diesel generator air start solenoids would function as required to mitigate an accident. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance

Enforcement.

The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," which states in part: "Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions." Contrary to the above, the licensee failed to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, as of July 22, 2011, the licensee failed to evaluate that the voltage available at the emergency diesel generator diesel engine air start solenoid would be sufficient to ensure starting the diesel engine under design basis conditions. The licensee failed to incorporate the required minimum operating voltage for the emergency diesel generator air start solenoids into Calculation E4C-017. This finding was entered into the licensee's corrective action program as Nuclear Notifications NN-201513266 and NN-201566686. Because this violation was of very low significance (Green) and has been entered into the licensee's corrective action program, this violation is being treated as a noncited violation consistent with the NRC Enforcement Policy: NCV 05000361/2011010-02, "Failure to Evaluate that Sufficient Voltage Would be Available at the Emergency Diesel Generator Air Start Solenoid."

.2.14 480 Volt Load Center Transformer 3B06X

a. Inspection Scope

The team reviewed the Updated Final Safety Analysis Report, system design criteria, selected drawings, and operating procedures for 480 Volt Load Center Transformer 3B06X. The team also performed walkdowns, reviewed Nuclear Notifications (condition reports) for recurrence of adverse conditions affecting reliability, and held discussions with cognizant licensee individuals. Specifically, the team reviewed:

• Updated Final Safety Analysis Report for the change (increase) in load center transformer rating

• One line diagrams for transformer 3B06X and load center 3B06

• Transformer and load center vendor nameplate data

• Engineering Change Package (ECP) 040201281-2 for installation of 3B06X cooling fans

• Calculation for transformer and load center bus overcurrent protection settings

• Load flow calculation for design basis transformer loading

• Surveillance test of transformer overcurrent relays

• Operating Instruction and Emergency Operating Instruction for load restrictions on load center 3B06

b. Findings

Failure to Incorporate Design Requirements into Procedures and Instructions

Introduction.

The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," because the licensee did not incorporate the design basis requirement from the vendor nameplate for maximum allowable amperage for load center 3B06, identified and required by Engineering Change Package ECP 040201281-2, in Operating Instruction SO23-6-3.

Description.

The team reviewed Updated Final Safety Analysis Report subsection 8.3.1.1.3.9, Class 1E Equipment Capacities, part B. 480 Volt Load Centers, 1. Transformer, which showed the 3B06X transformer rating as 1500/2000 kVA, AA/FA. The team also reviewed engineering change package ECP 040201281-2 for installation of load center transformer 3B06X cooling fans, which extended and increased the transformer rating from 1500 kVa AA to 1500/2000 kVa AA/FA. The engineering change package stated in the Description of Change, that, "The rated (FA) current capability delivered by the Transformer will exceed the 2000 amperage rating of the 480 volt Circuit Breaker and bus bars of the Load center, therefore, the load on the transformer shall be limited to 1663 kVa (2000 amperes). This restriction will also be added in 480 Volt Switchgear Operating Instruction SO23-6-3." The team reviewed plant Operating Instruction SO23-6-3 and determined that the load restriction was not incorporated as required by the engineering change package. The team also reviewed Emergency Operating Instruction SO23-12-11 attachments for cross-connecting Class 1E 480 volt buses between units. Operating Instruction SO23-12-11, Attachment 23, can be utilized to cross-connect Load Center Bus 3B06 to Load Center Bus 2B06 and supply 480 volt power to bus 2B06. The team found that the instruction did not restrict Load Center Bus 3B06 loading to 1663 kVa (or 2000 amperes) during the bus cross-tie condition in order to maintain the load center within the design basis rating. The team determined that due to the lack of information in operating instructions to limit load, plant operators may not maintain Load Center Bus 3B06 within the maximum current rating during plant upset conditions, which could result in the potential for load center damage or failure.

Analysis.

The team determined that the failure to incorporate the vendor required amperage limit, identified in engineering change package ECP-040201281-2, for bus load limit requirements, into 480 Volt Switchgear Operating Instruction SO23-6-3 was a performance deficiency. The finding was more that minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. In accordance with Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the finding was determined to have very low safety significance (Green) because it did not represent a design issue resulting in the loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. Specifically, the licensee had never implemented 480 Volt Switchgear Operating Instruction SO23-6-3 for the purpose of cross tying busses in an emergency, where the limiting load on the bus may have been exceeded. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.

Enforcement.

The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, "Design Control," which states in part: "Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions."

Contrary to the above, the licensee failed to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, as of July 22, 2011, the licensee did not incorporate the vendor required amperage limit, identified in engineering change package ECP-040201281-2, for bus load limit requirements, into 480 Volt Switchgear Operating Instruction SO23-6-3. This finding was entered into the licensee's corrective action program as Nuclear Notification NN-201570846. Because this violation was of very low significance (Green) and has been entered into the licensee's corrective action program, this violation is being treated as a noncited violation consistent with the NRC Enforcement Policy: NCV 05000361/2011010-03, "Failure to Incorporate Design Requirements into Procedures and Instructions."

.2.15 Auxiliary Feedwater Pump 2P-141 Discharge Flow Control Valve 2HV4713 Motor Starter

a. Inspection Scope

The team reviewed the Updated Final Safety Analysis Report, system design criteria, selected drawings, and operating procedures for discharge flow control valve 2HV4713 motor starter. The team also performed walkdowns, reviewed Nuclear Notifications (condition reports) for recurrence of adverse conditions affecting reliability, and held discussions with cognizant licensee individuals. Specifically, the team reviewed:

• Schematics and wiring diagrams for valve 2HV4713 motor starter

• Calculation for minimum voltage available at motor terminals during design basis degraded voltage conditions

• Calculation for minimum voltage at motor starter contactor coils during design basis degraded voltage conditions

• Preventive maintenance performed on 2HV4713 motor starter and Limitorque motor operator

b. Findings

Failure to Evaluate the Effects of Fuse Resistance and High Energy Line Break Conditions on Control Circuit Voltage

Introduction.

The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," because the licensee failed to incorporate into Design Calculations E4C-084 and E4C-085 the control power transformer circuit fuse resistance, including fuse clip resistance, and the temperature effects on cable resistance due to Auxiliary Feedwater Building area High Energy Line Break.

Description.

The team reviewed Calculation E4C-084, Unit 2 MCC Control Circuit Voltage Analysis, and found the calculation failed to consider the resistive load of the circuit fuses and fuse clips on the control power transformer control circuit, which were assumed to have negligible resistance. When questioned by the team, the licensee measured control power fuses, which included the fuse holder resistance. The licensee confirmed by measurement that the 1 ampere and 2 ampere fuse in series in the 2HV4714 contactor control circuit added approximately 2.5 ohms resistance to the control circuit. The team considered the additional 2.5 ohms resistance to be significant relative to the total resistance of the control circuit cable and contacts, which was approximately 10.3 ohms. Also, during a plant walkdown, the team identified that the licensee's calculation had also not considered the effect on control circuit cable resistance due to a high energy line break conditions in the auxiliary feedwater building when they determined the cable resistance in the control circuit voltage analysis. A preliminary analysis by the licensee during the inspection found that adequate voltage was available for the 2HV4713 contactor when the above noted deficiencies were corrected.

However, in reviewing the extent of condition for the fuse resistance deficiency, the licensee performed interim calculations which determined that 15 additional contactor control circuits would not meet the minimum voltage acceptance criteria based on the assumed conservative value for the control power transformer turns ratio. The licensee performed an additional interim calculation that took credit for a less conservative control power transformer turns ratio, but which was considered to be representative of the installed control power transformers. The less conservative control power transformer turns ratio, which was assumed based on the specifications for replacement control power transformers, improved the calculated voltage available to an acceptable value, which the licensee considered to provide a reasonable basis for immediate operability.

Subsequent to the inspection, the licensee verified the actual, installed, control power transformer turns ratio by field tests, and found the turns ratio to be acceptable for the 15 control circuits that were initially found to not meet the minimum voltage analytical limit for contactor operation.

Analysis.

The team determined that the failure to fully evaluate the circuit load in determining design limits in electrical calculations for degraded voltage conditions was a performance deficiency. The finding was more than minor because it was associated with the mitigating systems cornerstone attribute of design control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the team determined that the finding was of very low safety significance (Green) because it did not represent a design issue resulting in the loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. Specifically, the licensee performed subsequent preliminary analyses which demonstrated that the control circuits, where marginal voltage was available, would function as required to mitigate an accident. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.

Enforcement.

The team identified a Green noncited violation of 10 CFR 50, Appendix B, Criterion III, "Design Control," which states in part: "Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions." Contrary to the above, the licensee failed to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, as of July 22, 2011, the licensee failed to incorporate the fuse resistance, fuse clips resistance, and cable temperature and resistance effects (for Auxiliary Feedwater High Energy Line Breaks only), into Calculations E4C-084 and E4C-085, for degraded voltage conditions. This finding was entered into the licensee's corrective action program as Nuclear Notification NN-201546570 and NN-201550186.

Because this violation was of very low significance (Green) and has been entered into the licensee's corrective action program, this violation is being treated as a noncited violation consistent with the NRC Enforcement Policy: NCV 05000362/2011010-04, "Failure to Evaluate the Effects of Fuse Resistance and High Energy Line Break Conditions on Control Circuit Voltage."

.2.16 Safety Related Instrumentation Inverter 3Y002

a. Inspection Scope

The team reviewed the Updated Final Safety Analysis Report, system design criteria, selected drawings, and operating procedures for safety related instrument inverter 3Y002. The team also performed walkdowns, and held discussions with cognizant licensee individuals. Specifically, the team reviewed:

• The inverter capacitor replacement program

• 120 Vac vital instrument power one-lines

• Vendor manuals and supplements for the 125 Vac inverter

• Vendor schematics, wiring diagrams, and layout drawings for the inverter

• Maintenance activities to ensure the components are being maintained in accordance with vendor recommendations

• Calculations of record that determine and identify the loading margins for design basis conditions

• Calculations of record that determine and identify minimum and maximum operating temperature limits, minimum and maximum voltages at the terminals of the inverters for design basis conditions

• Last 3 system health reports covering these inverters; include supporting condition reports and any associated operability evaluations and root / apparent cause evaluations

• Summary listing of condition reports associated with all vital inverters

• Summary listing of replacement history for these inverters (starting with original equipment) and any design modifications associated with these inverters

• The last three completed surveillance and preventative maintenance procedures performed

b. Findings

No findings were identified.

.2.17 Reserve Auxiliary Transformer 2XR1 Fans, Pumps, and Controls

a. Inspection Scope

The team reviewed the Updated Final Safety Analysis Report, system design criteria, selected drawings, and operating procedures for Reserve Auxiliary Transformer 2XR1 fans, pumps, and controls. The team also performed walkdowns, and held discussions with cognizant licensee individuals. Specifically, the team reviewed:

• Calculation and vendor data sheets for auxiliary device settings

• Drawings, single line and distribution panel, showing plant electrical power feeds to the transformer control panel, pumps, and fans, including feeder cable sizes

• Vendor data - transformer tests, bushing tests, and surge arrestor tests

• Vendor schematics, wiring diagrams, and layout drawings for transformer components, controls and auxiliaries

• Condition monitoring procedure/requirements and results/trend data (since unit installation) for transformer, bushings, and surge arrestors

• Maintenance activities to ensure the components are being maintained in accordance with vendor recommendations

• The last PM/calibration test on control power feeder breaker(s) to transformer auxiliaries

b. Findings

No findings were identified.

.2.18 Class1E 600 Volt Cable

a. Inspection Scope

The team reviewed the Updated Final Safety Analysis Report, system design criteria, selected drawings, and operating procedures for Class 1E 600 Volt Cable. The team also performed walkdowns, and held discussions with cognizant licensee individuals. Specifically, the team reviewed:

• Latest Calculation(s) for Class 1E 600 Volt Power Cable Ampacity including any scheduled changes

• Latest Calculation(s) for 480 Volt Power Circuit Breaker settings

• Copies of modifications, on 480 Volt Load Centers or 480 Volt Motor Control Centers, within the last 5 years include Modification Summary and Technical Evaluation sections only

• Maintenance activities to ensure the cables are being maintained in accordance with vendor recommendations

b. Findings

No findings were identified.

.3.1 Operating Experience Smart Sample FY2008-01 - Negative Trend and Recurring Events Involving Emergency Diesel Generators

a. Inspection Scope

NRC Operating Experience Smart Sample (OpESS) FY 2008-01 is directly related to NRC Information Notice (IN) 2007-27, "Recurring Events Involving Emergency Diesel Generator Operability." The team performed a detailed review of this operating experience item to verify that the licensee had appropriately assessed potential applicability to site equipment and initiated corrective actions where necessary. The team independently walked down the Unit 2 and Unit 3 emergency diesel generators on several occasions to inspect for indications of vibration-induced degradation on emergency diesel generator piping and tubing and for any type of leakage (e.g., air, fuel oil, lube oil). The team performed also held discussions with cognizant licensee individuals and reviewed corrective action documents. Specifically, the team reviewed:

• The licensee's evaluation of IN 2007-27 and associated corrective actions

• Emergency diesel generators system health reports, emergency diesel generator notifications and work orders, and surveillance test results to verify that the licensee appropriately dispositioned emergency diesel generator concerns

• The 3A emergency diesel generator monthly surveillance on July 5, 2011, and performed pre- and post-run emergency diesel generator walkdowns to ensure the licensee maintained appropriate configuration control and identified deficiencies at a low threshold

b. Findings

No findings were identified.

.3.2 Inspection of Generic Letter 2007-01 - Inaccessible or Underground Power Cable Failures that Disable Accident Mitigation Systems or Cause Plant Transients

a. Inspection Scope

The team reviewed the Updated Final Safety Analysis Report, system design criteria, selected drawings, and operating procedures for submergence of cable, including flood protection systems and the flood level sensors. The team also performed walkdowns, and held discussions with cognizant licensee individuals. Specifically, the team reviewed:

• Listing of manholes inspected, date inspected and any photos taken during inspections

• The design basis for monitoring water in the vaults (i.e., level alarms)

• List of the cable vaults/tunnels/manholes and a listing of the cables and level switches inside

• Maintenance activities to ensure the level switches are being maintained in accordance with vendor recommendations

• Engineering evaluations and engineering changes associated with the flood detection system

b. Findings

Failure to maintain equipment important to safety

Introduction.

The team identified a Green noncited violation of 10 CFR Part 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants", in that appropriate corrective actions were not taken when the performance or condition of structures, systems, or components did not meet established goals. Specifically, the licensee failed to adequately monitor the condition of the Flood Level Detecting system in a manner to provide reasonable assurance the system could perform its intended function. The licensee also failed to properly evaluate Maintenance Rule Functional Failures and take appropriate corrective actions to improve system performance.

Description.

The Flood Protection System at the San Onofre Nuclear Generating Station consists of both passive and active components. The passive components are culverts, surface grading, subsurface drains, watertight doors, and the seawall. The passive components minimize flooding damage due to excessive rainfall, high tides, and wind generated waves. These passive components also handle seismic generated waves and ruptures of pipe and tanks. The active components of the Flood Protection System are the flood level sensors which are comprised of a: 1) float water level sensor, 2) switch, and 3) the associated alarm system. When area water levels begin to rise, these floor mounted devices actuate and provide annunciation at control room panels 2(3) CR57 to warn the control room operators, and local indication at panels 2(3) ZL9480-1 and or panels 2(3) ZL9481-2, that local flooding has exceeded the ability of the normal sump system.

There are 92 flood level detector sensors installed in separate areas and rooms located throughout the plant in both units. The licensee's Maintenance Rule Function Report MR-INST-01 identifies 37 different areas as having safety related flood level switches whose functional failures should be monitored. The Flood Protection subsystem is in the scope of 10 CFR 50.65, Maintenance Rule, because the flood sensors are safety-related (Quality Class II) as well as its function to mitigate accidents. The arrangement is that a single room location associated with a specific Train may have one or two flooding sensors. In the case where there is one sensor in a room, a single component failure would render the flooding detection system for that Train inoperable. Rooms having two sensors require both sensors to fail before flooding detection for that Train would be considered inoperable. Due to several documented failures the licensee has placed these components in 10 CFR 50.65(a)(1) status since September 2005.

There have been several modes of failure documented for the flood level sensors including: corrosion, stuck floats, switch mechanisms stuck, both float and switch mechanism stuck, moisture intrusion, and switches that would not activate when the float was in close proximity to the switch. Maintenance personnel have been able to free up the float and or switch mechanisms by cleaning, lubricating and exercising the mechanisms. From June 23, 2005 to January 16, 2007, five flood level sensors failed their respective surveillances and required cleaning and exercising in order to pass the surveillance test.

On January 23, 2006 the licensee's Maintenance Rule Expert Panel directed the Action Request Committee (ARC) to perform an operability review of these failed switches. The Action Request (corrective action document) which identified switches that had failed on June 23, 2005 and July 1, 2005, described a degraded condition where the equipment that had failed to operate on the initial test, were cleaned and lubricated, and the mechanisms exercised, and then functioned correctly. The Action Request did not identify the cause of the initial failure or actions to correct future problems. Therefore, the ARC concluded that the functional capability of the equipment was in question and it may not perform its intended function when required. The licensee concluded since the flood detectors failed the initial surveillance test and required pre-conditioning to yield acceptable test results that the switches identified in the Action Request were considered inoperable. The licensee set interim goals to 1) determine the cause of the failures, 2) replace detectors that had previously failed, and 3) sample a population of flood detectors to identify the possibility of a generic failure mechanism. The licensee also developed a plan to test the degraded flood level sensors every six months and required that the degraded flood level sensors must pass the tests without pre-conditioning every six months until appropriate corrective action was taken. On February 3, 2006 the first re-test was performed on the previously four failed units with following results: two of the switches passed, and two switches failed. Due to the subsequent failure of two of the units, the licensee took immediate action to replace both units and to continue the six month surveillance interval for the two units that had passed. These two units continued to be monitored every six months until they were replaced on January 25, 2008. On January 16, 2007, 3LSH9472-2, Safety Equipment Building Piping Tunnel Flood Detector Level failed its respective surveillance test and was placed into the same six month monitoring program until it was replaced on January 21, 2008. On March 28, 2008 the Operability Assessment that required the six month monitoring interval was closed since all the units that were in the program had been replaced.

During the same time frame four additional switches failed. Two of the switches were replaced, but due to lack of replacement parts, Flood Level Detectors 2LSH9500-1, Diesel Generator Building Piping Trench Area Flood Detector Level, and 2LSH9462-1, Charging Pump P190 Area Flood Detector Level, were documented as being mechanically exercised, but were not placed in the accelerated monitoring program. The team identified that on April 25, 2009, another flood level sensor, 2LSH9477-2, failed its surveillance test, was cleaned and exercised and placed back in service without being placed in an accelerated monitoring program. The failure and preconditioning was documented in NN-200404926 without identifying any other compensatory actions. As of July 22, 2011, all three of these flood level detectors had not been replaced or looked at since being placed back in service on or before April 25, 2009, and not placed into an accelerated inspection plan.

The team also identified that in the licensee's corrective action document AR-05070064-01, page 2 of 15 in the summary of results section, that the Flood and Sensor Alarm System did not meet its Maintenance Rule functional failure Performance Criterion Exceedence value of two per unit over a 36-month period ending July 2005. During this period, four flood sensors failed to actuate. Two of the failures together constitute a single maintenance rule functional failures, and the other two failures were individual maintenance rule functional failures. Thus the total number of maintenance rule functional failures for this 36 month period was three. On June 28, 2011, the Maintenance Rule Expert Panel decided to revise the performance criteria and changed it from an exceedence value of two functional failures to an exceedence value of three functional failures or 1 repeat (same root cause) functional failure over a 36-month period. This was documented in licensee document MR-INST-01. The team further identified in MR-INST-01 that the establishment of three functional failures was based upon a review of the Maintenance Orders for the period April 1, 1998 to March 31, 2011 and a review of Nuclear Notifications (corrective action documents) found three functional failures per unit per 36 month monitoring period. This would have been acceptable with the new, revised exceedence value of three, but the team identified that one of the failures identified for Unit 2 was actually a Unit 3 failure and that seven additional Unit 2 failures were not identified and three additional Unit 3 failures were not identified. Also, six of these identified failed flood level switches were not captured by Maintenance Work Orders or Nuclear Notification (corrective action documents). These failed switches should have been identified as Maintenance Rule Functional Failures and appropriate corrective actions taken.

The team also identified that five flood level sensors appeared to be beyond their required three year surveillance interval plus a 25 percent extension (3 years and 9 months). The team reviewed written information that the surveillance interval for these switches had been extended for a period not to exceed four years, without any conclusive justification, as documented in Nuclear Notification NN-200822111. Currently these five level switches are beyond the four year inspection interval. The extension of the surveillance places these units beyond their calculated calibration drift surveillance interval of once every four years,. The licensee further asserts that these particular switches are scheduled to be replaced within the next year. The team determined that the surveillance interval had been extended even though these types of switches have had a history of frequent failures. The team did not identify any additional compensatory measures or justifications for the five identified switches where their surveillance frequency had now been extended to 4 years,

Engineering has decided to conservatively replace all flood level switches as they have concluded that the failures appear to be related to aging of the sensors. The plan is to replace all the switches between two and four years during their scheduled preventative maintenance activities.

Analysis.

The team determined that the failure to properly maintain the flood level sensors which are used for control room annunciation to warn the control room of flooding of a space that has safety related or important to safety components, was a performance deficiency. The finding was more than minor because it was associated with the mitigating systems cornerstone attribute of design control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not maintain flood level sensors appropriately to provide reasonable assurance that the components would be capable of fulfilling their intended function. In accordance with Inspection Manual Chapter 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," the team determined that the finding represented the degradation of equipment and functions specifically designed to provide notification to the control room of flooding of spaces with safety related or important to safety equipment and components. Therefore, the finding was potentially risk significant and a Phase 3 analysis was required. The preliminary significance determination was based on Inspection Manual Chapter 0609, Appendix M, "Significance Determination Process Using Qualitative Criteria." The senior reactor analyst determined qualitatively that the risk was very low for the following reasons: (1) the frequency of internal flooding is very low, (2) floods in most of the problem areas would not cause a transient, (3) redundant indications of flooding exist, including fire and sump pump operations, and (4) none of the potentially flooded areas would likely affect more than one train of safety equipment.

This finding involved a cross-cutting aspect in the area of Human Performance, Resources, because the licensee failed to assure that equipment and other resources were available and adequate to assure nuclear safety. Specifically, the licensee was not able to maintain the flood level switches adequately to assure nuclear safety due to long-standing equipment issues. H.2(a).

Enforcement.

The team identified a Green noncited violation of 10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," which states in part: "Each holder of a license to operate a nuclear power plant shall monitor the performance or condition of structures, systems, or components, against licensee-established goals, in a manner sufficient to provide reasonable assurance that such structures, systems, and components, are capable of fulfilling their intended functions, and when the performance or condition of a system, structure, or component, does not meet established goals, appropriate corrective actions shall be taken." Contrary to the above, the licensee failed to take appropriate corrective actions when the performance or condition of a system, structure, or component, did not meet established goals. Specifically, as of July 22, 2011, the licensee failed to adequately monitor the condition of the Flood Level Detecting system in a manner to provide reasonable assurance the system could perform its intended function. The licensee failed to properly evaluate Maintenance Rule Functional Failures and take appropriate corrective actions to improve system performance. These level switches are connected to control room annunciation to warn the control room of flooding in a space that has safety-related or important to safety components. This has been entered into the licensee's corrective action program as Nuclear Notifications NN-201567315 and NN-201570575. Because this violation is of very low safety significance and has been entered into the licensee's corrective action program, this violation is being treated as a non-cited violation consistent with the NRC Enforcement Policy: NCV 05000361/2011010-05, "Failure to Maintain Equipment Important to Safety."

.4 Results of Reviews for Operator Actions:

The team selected risk-significant components and operator actions for review using information contained in the licensee's probabilistic risk assessment. 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

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 selected operator actions were:

• Loss of component cooling water to the Reactor Coolant Pumps seals due to a loss of instrument air system

• Loss of Offsite Power with Emergency Diesel Generator malfunctions requiring unloading of vital DC buses and establishing ventilation of the Engineered Safety Features Switchgear rooms

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems

a. Inspection Scope

The team reviewed actions requests associated with the selected components, operator actions and operating experience notifications.

b. Findings

Failure to Adequately Control Document Changes

Introduction.

The team identified multiple examples of a Green non-cited violation of 10 CFR 50, Appendix B, Criterion VI, "Document Control," because the team identified numerous drawing inconsistencies where changes to certain components were not changed on all affected drawings, and procedural errors where changes were not made to all affected documents.

Description.

The team started the inspection with one week of in-office preparation, reviewing documentation related to the components selected, and continued this review when the team arrived on site. During the review of drawings, procedures, and calculations, the team noted numerous errors within the documents. References to incorrect sections in procedures, drawings that did not reflect changes due to modifications, drawings that were not accurate from one drawing to another drawing depicting the same components, and inconsistent component requirements between vendor recommendations and stress analysis calculations, were noted. Within the first week of the inspection the team had noted at least nine examples of documentation errors.

Example 1 On June 23, 2011 while reviewing wiring diagram SO23-302-4-2-268, the team identified a white indicating light found on the diagram was not included on elementary diagram 30956 Sheet 1. The licensee further identified that the white indicating light was also omitted from control circuit loading calculation E4C-084. Nuclear Notification NN-201511720 Example 2 Incorrectly marked "NO" box with respect to a change impacting PRA with respect to removal of the level switch low-low trip contact for emergency diesel generator fuel transfer pump. Nuclear Notification NN-201510143 Example 3 A modification removed diesel fuel storage tank low-low level bi-stables. Figure III-1 on page 135 still shows a "/Y1" symbol, where the LSLL5903 symbol should only contain the "X1" suffix. Nuclear Notification NN-201510265 Example 4 Current revision 13 of One Line Drawing 30127 shows the cable size from 2B1305 to MCC 2BC as 3-1/C 350 MCM. Current Revision 23 of One Line Drawing 30135 shows the same cable as 3-1/C500 MCM. Nuclear Notification NN-201512987 Example 5 Procedure SO23-V-3.4 page 121 of 126 on attachment 9, the formula for "LnRWST Temp" and "Ln_RCS_Temp" are identical. The formulas for the two values should be different and the formulas should reflect it. Nuclear Notification NN-201513110 Example 6 Emergency Diesel Generator air start solenoid air valves 2(3)HY5955A1, B1,C1, D1,A2, B2, C2, D2, are incorrectly shown on drawing SO23-403-12-44 and in the associated SAP Master Data functional location records. Nuclear Notification NN-201513112 Example 7 Section 8.5.10 of current revision 15 of electrical calculation E4C-042 points to an incorrect reference. Reference 6.24 should be reference 6.23. Nuclear Notification NN-201513415 Example 8 Main One Line Drawing 30101 only shows the lower rating of 1500 kVA for transformer 3B06X, whereas One Line Drawing 32120 shows both the lower 1500 kVA and higher 2000 kVA rating for transformer 3B06X. Nuclear Notification NN-201546498 Example 9 CCW Surge Tank Vendor Drawing # 5782 calls out anchor bolt material to be SA-145, Grade B7, whereas stress analysis calculation C-259, Section 2.03.04, Revision 0 calls out A-307 for the bolt material. Nuclear Notification NN-201571137

Individually, the errors were not significant. Collectively, these document errors demonstrated a weakness in the licensee's program for review and approval of controlled documents that were used in the plant. Other errors were also noted throughout the inspection, and were captured by the licensee in their corrective action program.

Analysis.

The team identified that collectively, from a program perspective, the failure to properly incorporate design changes of components in the plant to all affected drawings, procedures, or instructions, was a performance deficiency. The finding was more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. In accordance with Inspection Manual Chapter 0609, 4, "Phase 1 - Initial Screening and Characterization of Findings," the team determined that the finding was of very low safety significance (Green) because it did not represent a design issue resulting in the loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. Specifically, none of the documents with the identified errors had been used in response to any events or plant perturbations. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance

Enforcement.

The team identified a Green non-cited violation with multiple examples of 10 CFR 50, Appendix B, Criterion VI, "Document Control," which states in part: "Measures shall be established to control the issuance of documents, such as instructions, procedures, and drawings, including changes thereto, which prescribe all activities affecting quality. These measures shall assure that documents, including changes, are reviewed for adequacy and approved for release." Contrary to the above, the licensee failed to assure that documents, including changes, were reviewed for adequacy and approved for release. Specifically, on June 23, 2011, the team identified numerous drawing inconsistencies where changes to certain components were not changed on all affected drawings and procedural errors where changes were not made to all affected documents. The licensee has entered the errors into their corrective action program under numerous Nuclear Notifications listed above. Because this violation is of very low safety significance and has been entered into the licensee's corrective action program, this violation is being treated as a non-cited violation consistent with the NRC Enforcement Policy: NCV 05000361,05000362/2011010-06, "Failure to Adequately Control Document Changes."

OA6 Meetings, Including Exit

On July 22, 2011, the team leader presented the preliminary inspection results to Mr. P. Dietrich, Senior Vice President and Chief Nuclear Officer, and other members of the licensee's staff. On September 13, 2011, the team leader conducted a telephonic final exit meeting with Mr. Mr. R. St. Onge, Director, Nuclear Regulatory Affairs, and other members of the licensee's staff. The licensee acknowledged the findings during each meeting. While some proprietary information was reviewed during this inspection, no proprietary information was included in this report.

4OA7 Licensee Identified Violations

No findings were identified.

s: 1 - Supplemental Information ATTACHMENT 1

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

D. Axline, Project Manager, Nuclear Regulatory Affairs

D. Bauder, VP & Station Manager

M. Carr, Manager, NFM/PRA

J. Dahl, Manager, Operations

P. Dietrich, SVP & Chief Nuclear Officer

S. Dornseif, Technical Specialist, Nuclear Regulatory Affairs

M. Farmer, Manager, Health Physics

J. Hays, Engineer, DEO Mechanical

K. Johnson, Manager, Design Engineering

G. Kline, Sr Director, Engineering

J. Kolons, Engineer, Nuclear Regulatory Affairs

D. LeGare, Engineer, Plant Engineering

J. Madigan, Director, Nuclear Safety Culture and Site Recovery

L. McCann, Manager, Chemistry

T. McCool, Plant Manager

R. McWey, Manager, Oversight

C. Miller, Manager, Operations

V. Nazareth, Supervisor, NFM

D. Nougier, Engineer, DEO Mechanical

R. Pettus, Technical Specialist, Nuclear Regulatory Affairs

L. Rafner, Engineer, Nuclear Regulatory Affairs

C. Robinson, Engineer, Nuclear Regulatory Affairs

P. Schofield, Engineer, DEO Mechanical

B. Sholler, Director, Maintenance & Construction

J. Sills, Project Manager, Performance Improvement

R. St. Onge, Director, Nuclear Regulatory Affairs

R. Trapp, Engineer, DEO Mechanical

R. Treadway, Manager, Nuclear Regulatory Affairs

S. Atkins, Engineer, DEO Electrical

G. Hansen, Engineer, IST Program

E. Mensa-Wood, Plant Engineer

G. Segich, Plant Engineer

R. Taylor, Plant Engineer

D. Tuttle, Supervisor, EDG Team

R. Urena, Program Engineer

T. Yackle, Assistant Plant Manager

NRC personnel

S. Achen, Resident Inspector

G. Warnick, Senior Resident Inspector,

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000361/2011010-01

NCV Inadequate Assessment of Seismic Qualification of Emergency Diesel Generator Buried Fuel Oil Tanks.

(1R21.2.9)

05000361/2011010-02

NCV Failure to Evaluate that Sufficient Voltage would be Available at the Emergency Diesel Generator Air Start Solenoid. (1R21.2.13)

05000361/2011010-03

NCV Failure to Incorporate Design Requirements into Procedures and Instructions. (1R21.2.14)

05000362/2011010-04

NCV Failure to Evaluate the Effects of Fuse Resistance and High Energy Line Break Conditions on Control

Circuit Voltage. (1R21.2.15)

05000361/2011010-05

NCV Failure to Maintain Equipment Important to Safety. (1R21.3.2)

05000361;

05000362/2011010-06

NCV Failure to Adequately Control Document Changes. (4OA2)

LIST OF DOCUMENTS REVIEWED

Calculations

NUMBER TITLEREVISION / DATE

A-95-NM-MOV-PL/TB

GL 95-07 Pressure Locking and Thermal Binding Screening Evaluation

2A-98-E-001

Non-1E Interactions in Safety Related Process Systems

1C-259-02.03.04

Component Cooling Water SurgeTanks Seismic Loads

2E4C-031 Cable sizing to accommodate available short circuit currents

2E4C-042 8KV Cable Ampacity

15E4C-051.1 Class 1e 600 volt power cable ampacity for 480 volt load center feeders

1E4C-084 Unit 2 MCC Control Circuit Voltage Analysis

0E4C-086 Songs 2 & 3 data development and documentation

7E4C-090 Auxiliary System Voltage Regulation

7E4C-098 4KV Switchgear Protective Relay Setting Calculation

3E4C-099 Sr 480 volt power circuit breaker settings

3E4C-102

GL 89-10 MOV Voltages During Design Basis Accidents

Calculations

NUMBER TITLEREVISION / DATE

E4C-102 (ECN

D0003955)

GL 89-10 MOV Voltages during Design Basis Accident

3E4C-105 Electrical Heat Load Eval. For SBO

3E4C-116 Non-1E 480 volt MMC protection calculation

4E4C-120 600 volt power cable ampacity for 480 volt MMC & 120 volt ac panels 4J-BHA-001 Safety Injection System SDC Temp Indicators TLU Calculation

1J-BHA-081 Assessment of Shutdown Cooling System Flow Indication

1J-BHA-111 LPSI Pump(S) Suction And Discharge Pressure Indicator Uncertainty

0J-EPA-002 TLU For Saltwater Flow To CCW Heat Exchangers 2(3) E001a And 2(3)002b

J-JEA-002 Fuel Oil Level Setpoints for EDG Storage tanks

J-JEA-002 Fuel Oil Level Setpoints for EDG Storage Tanks

2J-RNA-015 Various room flooding alarm sensor setpoint calculation

1M-0012-002

Safeguards Pumps with Suction from Refueling Water Tank December 8, 1976

M-0012-01D

NPSH of ESF Pumps

3M-0012-027

LPSI IST Minimum Performance Requirements

1M-0016-005

EDG Fuel Oil Tank Sizing

0M-0027-017

Safety-Related Backup Supply for the CCW Surge Tank

0M-0027-020

CCW Safety-Related Makup System Hydraulic Calculation

0M-0027-023 CCW/SWC heat exchanger operability 0

M-00783-116

Room Temperature Response During Station Blackout

2M-26.3 CCW Surge Tank Pressure

2M-26.4 CCW Surge Tank Sizing and Critical Crack Loss

3M-73-62 Elevation 50 foot esf switchgear rooms normal equipment sizing calcs October 22, 1974 M-76-40 EDG Building Heat Load Calculation

2M-76-41 EDG Fan Sizing Calculation

Calculations

NUMBER TITLEREVISION / DATE

M-76-42 EDG Duct Sizing Calculation

0M-8910-1204-OB-006

GL 89-10 Operational Basis Calculation for the Containment Emergency Sump ECCS Suction Valves

0M-8910-SP-

2HV9 301

GL 89-10 Setpoint Calculation for 2HV9301

1M-8910-SP-

2HV9 302

GL 89-10 Setpoint Calculation: Containment Sump Isolation Valve 2M-8910-WALT-WL-043

GL 89-10 Weak Link Evaluation of MOV Group No. 43

0SO2 3-407-4-28

CCW Surge Tank Seismic Analysis

4SO2 3-407-7-9

EDG Fuel Oil Tank Seismic Analysis

0S-PEC-10 Sizing of HP & LP Safety Injection PumpsAugust 7, 1970

S-PEC-355 SIS: Instrumentation Tolerances/LPSI System Test Acceptance Criteria Design Basis Documents

NUMBER TITLEREVISION / DATE DBD-SO23-120

6.9 kV, 4.16 kV & 480 V Electrical Systems

7DBD-SO23-140

Class 1E 125 Vdc System

8DBD-SO23-750

Emergency Diesel Generators

4DBD-SO23-780

Auxiliary Feedwater System

Drawings

NUMBER TITLE REVISION / DATE

002678 Sectional drawing heat exchanger, (Layout of RCPs) 0 16720 4 IN CL900 'Y' Globe Lift Check Valve 1 20195 Offshore Intake and Discharge Box- Sections and Details 4 20447 Sh. 3 Duo-Chek Valve

June 7, 1976

Drawings

NUMBER TITLE REVISION / DATE

21050 Sht 1 Electrical manholes 17 22002-0 (Bechtel Dwg. 10079003) Intake structure equipment arrangement sections 0 22030 Intake Structure General Arrangement-Isometric 2 22044 Offshore Intake and Discharge Box-Plan and Sections 6 23700 Sht 1 Electrical & piping underground galleries & access building floor plans

23856 Diesel Generator Building, Sections and Details 12 23865 Diesel Fuel Oil Storage Tank Enclosure Plan & Details 23 2LT5903-1 Diesel 2G002 Fuel Storage Tank Level 4/5 30094 One Line Diagram, 480 V LDCTR Device Function No. Relay Table & Legend, Sh. 1

30101 One line diagram main auxiliary power system 40 30101 One Line Diagram Main Auxiliary Power System 37 30103 One line diagram reserve auxiliary transformer protection 11 30105 One line diagram 6900 V switchgear bus 2A01 & 2A02 13 30106 One line diagram 4160 V switchgear bus 2A03 14 30107 One line diagram 4160 V switchgear bus 2A04 (ESF) 17

30109 One line diagram 4160 V switchgear bus 2A06 (ESF) 18 30109 One line diagram 4160 V switchgear bus 2A06 (ESF) 18 30110 One line diagram 4160 V switchgear bus 2A07 18 30111 One line diagram 4160 V switchgear bus 2A08 12 30112 One line diagram 4160 V switchgear bus 2A09 11

30113 One Line Diagram Diesel Generator Protection

6 30115 One line diagram 480 V loadcenter 2B01 10 30116 One line diagram 480 V loadcenter 2B02 11 30117 One line diagram 480 V loadcenter 2B03 20 30118 One line diagram 480 V loadcenter 2B04 (ESF) and 2B24 (ESF) 21 30120 One line diagram 480 V loadcenter 2B06 (esf) and 2B26 (ESF) 24

Drawings

NUMBER TITLE REVISION / DATE

30121 One line diagram 480 V loadcenter 2B07 19 30122 One line diagram 480 V loadcenter 2B08 19 30123 One line diagram 480 V loadcenter 2B09 21 30124 One line diagram 480 V loadcenter B10 17 30125 One line diagram 480 V loadcenter B11 20 30126 One line diagram 480 V loadcenter B112 23

30127 One line diagram 480 V loadcenter 2B13 13 30128 One line diagram 480 V loadcenter 2B14 18 30129 One line diagram 480 V loadcenter 2B15 14 30130 Sht 1 One line diagram 480 V loadcenter 2b16 18 30130 Sht 2 One line diagram 480 V loadcenter 2b18 9 30132 One line diagram 480 V motor control center 2BRA (ESF) 17 30134 One line diagram 480 V motor control center 2BRB (ESF) 17

30136 One line diagram 480 V motor control center 2BD (ESF) 19 30137 One line diagram 480 V motor control center 2BE (ESF) 38 30141 One line diagram 480 V motor control center 2BH (ESF) 23 30142 One line diagram, 480V motor control center 2BJ 31 30150 One line diagram 208/120 V lighting distribution bus l02 13

30151 One line diagram 208/120 V lighting distribution bus l01 12 30154 One line diagram 480 V motor control center BQ (ESF) 29 30156 One line diagram 480 V motor control center BS (ESF) 28 30161 One line diagram 480v motor control center 2BX 35 30162 One line diagram 480 V motor control center 2BY (ESF) 39

30164 One line diagram 480 V motor control center 2BZ (ESF) 43 30164 One line diagram, 480 V motor control center 2BZ 43 30166 One line diagram 480 V

MMC 208/120 Vac heater panels - (ESF) 48 30172 One line diagram class 1E 125 Vdc and 120 Vac power system 24 30173 One line diagram 125 Vdc switchboard 2D1 24

Drawings

NUMBER TITLE REVISION / DATE

30174 One line diagram 125 Vdc switchboard 2D1 25 30175 One line diagram 125 Vdc switchboard 2D3 17 30176 One line diagram 125 Vdc switchboard 2D4 20 30182 One line diagram 125 Vdc switchboard 2Y01 32 30183 One line diagram 125 Vdc switchboard 2Y02 30 30184 One line diagram 125 Vdc switchboard 2Y03 17

30185 One line diagram 125 Vdc switchboard 2Y04 15 30226 Elementary Diagram Electrical Auxiliary - 4.16 Kilovolt Bus 2A06 Tie Breaker (3A06)

30324 Diesel Generator Diesel Fuel Transfer Pump P0093 23 30327 Diesel Generator Diesel Fuel Transfer Pump P0095 21 30329 Sh. 1 Elementary Diagram Diesel Generator 2G002 Protection AC System,

30593 Elementary Diagram Reactor - CCW Return Line isolation valve TV9144

30682 Elementary Diagram, Refueling Water Tank Outlet

HV9301 10 30694 Elementary Diagram, Containment Emergency Sump Outlet HV9302

31353 Elementary diagram HVAC plant CB ESF switchgear room norm ac unit E430

31400 Elementary diagram engineered safety features switchgear ROPM exhaust fan a165

32106 Sht 1 One line diagram 4160 V switchgear bus 3A03 16 32107 One line diagram 4160 V switchgear bus 3A04 (ESF) 18 32109 One line diagram 4160 V switchgear bus 3A06 (ESF) 20 32110 One line diagram 4160 V switchgear bus 3A07 18 32111 One line diagram 4160 V switchgear bus 3A08 16

2112 One line diagram 4160 V switchgear bus 3A09 12 32115 One line diagram 480 V loadcenter 3B01 11 32116 One line diagram 480 V loadcenter 3B02 16 32117 One line diagram 480 V loadcenter 3B03 15

Drawings

NUMBER TITLE REVISION / DATE

32118 One line diagram 480 V loadcenter 3B04 (ESF) 23 32120 One line diagram, 480 V loadcenter 3B06 (ESF) 24 32121 One line diagram 480 V loadcenter 3B07 18 32122 One line diagram 480 V loadcenter 3B08 18 32123 One line diagram 480 V loadcenter 3B09 17 32125 One line diagram 480 V loadcenter 3B11 14

2126 One line diagram 480 V loadcenter 3B12 16 32127 One line diagram 480 V loadcenter 3B13 13 32128 One line diagram 480 V loadcenter 3B14 16 32129 One line diagram 480 V loadcenter 3B15 12 32130 Sht 1 One line diagram 480 V loadcenter 3B16 16 32130 Sht 2 One line diagram 480 V loadcenter 3b18 15 32136 One line diagram 480 V motor control center 3BD (ESF) 21

2137 One line diagram 480 V motor control center 3BE (ESF) 35 32141 One line diagram 480 V motor control center 3BH (ESF) 20 32142 One line diagram 480 V motor control center 3BJ (ESF) 28 32155 One line diagram 480 V motor control center 3BRA (ESF) 16 32155 Sht 2 One line diagram 480 V motor control center 3BRB (ESF) 16

2162 One line diagram 480 V motor control center 3BY (ESF) 36 32164 One line diagram 480 V motor control center 3BZ (ESF) 38 32166 One line diagram 480 V motor control center 208/120 V heater panels (ESF)

32172 One line diagram class 1e 125 Vdc & 120 Vac power system 17 32173 One line diagram 125 Vdc distribution switchboard 3D1 24 32174 One line diagram 125 Vdc distribution switchboard 3D2 23 32175 One line diagram 125 Vdc distribution switchboard 3D3 19 32176 One line diagram 125 Vdc distribution switchboard 3D4 17

2182 One line diagram 125 Vac vital bus 3Y01 23 32183 One line diagram 125 Vac vital bus 3Y02 23

Drawings

NUMBER TITLE REVISION / DATE

32184 One line diagram 125 Vac vital bus 3Y03 13 32185 One line diagram 125 Vac vital bus 3Y04 11 32226 Elementary Diagram Elec Aux - 4.16 Bus 3A06 Tie Breaker 2A06

32269 Elementary Diagram Electrical Auxiliaries Bus 3B06 Supply Breaker

3502 Sht 2 Underground electrical duct & manhole system explode view of MH.

35034 Sht1 Underground electrical DG. duct & manhole system exploded view of manhole

35035 Sht 2 Underground electrical DG. duct & manhole system exploded view of manhole

35036 Sht 3 Underground electrical DG. duct & manhole system exploded view of manhole

35063 Sht 4 Underground electrical DG. duct & manhole system exploded view of manhole

35119 Area 2C3 conduit & tray plan elevation (-) 2'-6" to elevation 11'-6"

35510 Area CA7 conduit and manhole plan below elevation 9'-0" 5 35511 Area CA8 conduit and manhole plan below elevation 9'-0" 6 35678 Area 3C3 conduit & tray plan elevation (-) 2'-6" to elevation 9'-6" 9 40110D Diesel Generator System (Train B) 33 40111A P&ID reactor coolant system NO. 1201 42 40111ASO3 P&ID reactor coolant system NO. 1201 36 40112A P&ID safety injection system NO. 1204 36 40112AS03 P&ID safety injection system NO. 1204 42

40112B P&ID safety injection system NO. 1204 36 40112BS03 P&ID safety injection system

NO 1204 38 40112CS93 P&ID safety injection system

NO 1204 31 40112DS03 P&ID safety injection system

NO 1204 25

- 10 -

Attachment Drawings NUMBER TITLE REVISION / DATE

40114BS03 P&ID containment spray system

NO 1206 16 40116A Diesel Fuel Storage System 13 40126BS03 P&ID component cooling water system (salt water pumps) 25 40127A CCW System (Pumps) 30 40127B CCW System (Tanks) 37 40127D CCW System (Supply Header) 17

40127FS03 P&ID Component Cooling Water System (non critical loop) 30 40127H CCW Back Up Nitrogen System 5 40127J Safety Related Make Up System 4 40141A Main Steam System 37 40160A Auxiliary Feedwater System 43 40177B Misc. Ventilation System (Turbine and Diesel Generator Building)

40816-2 Vendor Print: EDG Ductwork 2 41065 SH 4 Low pressure safety injection (LPSI) pump tag no. 3P016 IST curves

41108 SH 1 Salt water cooling pump 33WX-1 stage vertical circulator, (SWC Pump Details, Geometry)

5104346 Plan and Profile of Unit 2 Offshore Intake and Diffuser Outfall System

556-32919 Concentric series orifice plates (permutit) 4 5782 Vendor Print: CCW Surge Tank 6 C31100-043 Thermal Unit Assy. Square D Company J

D-52246 24-N-9126-EMO-S Gate Valve 5 F42902 24" Type 9220 Valve Assembly with Limitorque

SMB-00-10/H2BC Actuator

H FSAR Figure 10.4-2

Schematic elevation view, rooms and areas subject to flooding from failure of main condenser cooling water line maximum flood level

LOOP 2TE9144 Component Cooling Water From Reactor Cooling Pump P001 Seal Heat Exchanger Temperature

S023-405-33-Pump Performance Test Data, (SWC Pump P307) 0

- 11 -

Attachment Drawings NUMBER TITLE REVISION / DATE

M119 S05-2-3-2 Flow Orifice (V4105081) September 8,1975

S2-1204-ML-004 Sh. 1 Containment Spray Pump P-013 Suction from Containment Emergency Sump,

S2-SI-004-H-002 Pipe Support Assembly 2 SO23-302-15-2-1 Main Assy. 3B06X Transformer - ABB 0

SO23-302-4-2-

268 Connection Diagram 10 SO23-3-2-15.3 Nameplate 3B06X Transformer - ABB 2 SO23-3-2-15-1 Outline 3B06X Transformer - ABB 2

SO23-922-157 Byron jackson schematic diagram reactor coolant pump (4 mechanical seals) - (pages 336 - 349)

SO23-955-22 Wiring diagram inverter power stage 4 SO23-955-25 Cyberex Inc. Schematic, overload sensor relay 0 SO23-955-27 Cyberex Inc. Voltage feedback circuit (inverter) 1

SO23-955-28 Cyberex Inc. Sync. module without of/uf disconnect 1 SO23-955-29 Cyberex Inc. Schematic, modulation index control module 2 SO23-955-30 Cyberex inverter schematic, inverter logic power supply

1080 2 SO23-955-31 Cyberex Inc. Schematic, refference output 1 SO23-955-33 Cyberex Inc. Schematic, voltage sensor assembly 130

Vdc 1 SO23-955-51 Cyberex Inc. Assembly C1 and C2 non-PCB replacement 0 SO23-955-7 Cyberex Inc. Panel layout and component arrangement, inverter 2 SO23-955-8 Schematic, alarm circuit 4

- 12 -

Attachment Notifications

AR-010400750

NN-200309580

NN-200920504

NN-201300994

AR-011200343-1

NN-200314488

NN-200925876

NN-201353389

AR-020600675

NN-200317335

NN-200936736

NN-201355181

AR-020600778

NN-200365345

NN-200953896

NN-201355489

AR-041000476

NN-200404926

NN-200953928

NN-201358962

AR-050700064

NN-200405349

NN-200968445

NN-201359454

AR-060500265

NN-200408880

NN-200999308

NN-201360910

AR-070100845

NN-200413417

NN-201009885

NN-201382742

AR-070200583

NN-200421121

NN-201025656

NN-201388166

AR-070801423

NN-200458930

NN-201030569

NN-201393463

AR-070801527

NN-200475737

NN-201030615

NN-201464957

AR-070900135

NN-200491567

NN-201030697

NN-201409984

AR-980200847

NN-200492412

NN-201030953

NN-201419477

AR-980900844

NN-200493412

NN-201030958

NN-201427560

AR-990201844

NN-200525787

NN-201041585

NN-201440705

NN-200002218

NN-200528556

NN-201044387

NN-201455130

NN-200005159

NN-200531655

NN-201053506

NN-201464519

NN-200005383

NN-200533861

NN-201066362

NN-201477578

NN-200015510

NN-200548623

NN-201076929

NN-201502381

NN-200017657

NN-200557516

NN-201082650

NN-201510143*

NN-200017780

NN-200572793

NN-201092132

NN-201510265*

NN-200018233

NN-200581670

NN-201092157

NN-201510438*

NN-200022360

NN-200592570

NN-201092188

NN-201510529

NN-200024707

NN-200599255

NN-201103947

NN-201511720*

NN-200029561

NN-200063244

NN-201104752

NN-201512987*

NN-200110829

NN-200614996

NN-201105060

NN-201513110**

NN-200139401

NN-200615973

NN-201115449

NN-201514266*

NN-200139407

NN-200692347

NN-201115548

NN-201543871

NN-200139445

NN-200692658

NN-201122760

NN-201546498*

NN-200153831

NN-200695732

NN-201180567

NN-201546570*

NN-200161331

NN-200704606

NN-201217817

NN-201546801*

NN-200180393

NN-200711323

NN-201220380

NN-201548802

NN-200182897

NN-200711339

NN-201229835

NN-201550186*

NN-200191328

NN-200719514

NN-201241123

NN-201550697*

NN-200195248

NN-200722217

NN-201241496

NN-201555635

NN-200198219

NN-200738974

NN-201244687

NN-201566686*

NN-200200408

NN-200759986

NN-201245392

NN-201566717

NN-200211508

NN-200760074

NN-201249556

NN-201567315

NN-200214357

NN-200766430

NN-201250381

NN-201568826**

NN-200216613

NN-200771367

NN-201251951

NN-201569096*

NN-200256206

NN-200774219

NN-201255936

NN-201570619**

NN-200256262

NN-200780016

NN-201257122

NN-201570846*

NN-200256585

NN-200782241

NN-201257149

NN-201571139**

NN-200256588

NN-200811963

NN-201257150

NN-201572887*

NN-200258836

NN-200819851

NN-201257567

NN-201572888*

NN-200282342

NN-200822111

NN-201260709

NN-201572891*

NN-200282566

NN-200824442

NN-201260807

NN-201572903*

- 13 -

Attachment Notifications

NN-200282567

NN-200840027

NN-201260812

NN-201572905*

NN-200282568

NN-200904514

NN-201260817

NN-201573734*

NN-200293643

NN-200913309

NN-201260835

NN-200526477

NN-200919264

NN-201260876

Procedures

NAME TITLE REVISION / DATE

2PE-517-01 DG Building Normal and Emergency HVAC, Unit 2 1 SO123-I-1.9 Maintenance plan program 18 SO123-I-4.7 MCCB-inspection and Mechanical tests 13 SO123-I-4.7 MCCB - Inspection and Mechanical Tests 13

SO123-II-11.1 Test Procedure 8 SO123-II-11.152 Circuit Device Tests and Overall Functional Test 17 SO123-II-9.48 Magnetrol and other miscellaneous liquid level switches calibration

EC1 Permanent Plant Modifications

NUMBER TITLE REVISION/DATE

NECP NO. 8000072264

Replace non-1e cutler hammer buckets with starters in

MMC 2BC, 2BL, and 2BW

November 26, 2008 NECP NO.

800127363 Addition of 1E, seismic i, 750 kVa, transformer and load center to the existing 4.16 kV feeder supply breaker

3A0417 (unit 3 train A)

December 16, 2009 NECP NO.

800130452 Addition of 1E, seismic I, 750 kVa, transformer and load center to the existing 4.16 kV feeder supply breaker 2A0420 (unit 2 train A)

November 19, 2009 NECP NO.

800130487 Addition of 1E, seismic I, 750 kVa, transformer and load center to the existing 4.16 kV feeder supply breaker 2A0620 (unit 2 train B)

April 15, 2010 NECP NO. 8001559974

Replace inverter 3Y005 & battery charger 3B005 and install new battery charger 3B025

June 14, 2010 NECP NO.

800239532 NECP to replace unit 3, 480 volt load center breakers with

NLI masterpact breakers June 6, 2011

- 14 -

Attachment

Procedures

NAME TITLE REVISION / DATE

SO123-III-6.6.1 Diesel Fuel Oil Sampling and EDG Fuel Oil Storage Tank Chemical Additions

SO123-V-14 Oil sampling and analyses program 15 SO123-XV-1.20 Seismic Controls 2 SO23-12-11 EOI Supporting Attachments (14 - RAS Operation) 10 SO23-12-11 EOI Supporting Attachments 8

SO23-13-15 Loss of shutdown cooling, (abnormal operating instruction) 23 SO23-13-7 Loss of CCW/Salt Water Cooling 16 SO23-13-7 Loss Of Component Cooling Water (CCW)/Saltwater Cooling (SWC)

SO23-2-8

Salt Water Cooling System Operation 38 SO23-302-2-353

IB-8.2.7-2, Metal Clad Switchgear Instructions E SO23-3-2.6 Shutdown cooling system operation 29 SO23-3-2.7 Safety injection system operation 28 SO23-3-2.8.1 Refueling cavity draining operations 18 SO23-3-3.19 4kV Emergency Bus Transfer Test 17 SO23-3-3.23 Diesel Generator Monthly and Semi-Annual Testing 53

SO23-3-3.23.1 Diesel Generator Refueling Interval Tests 39 SO23-3-3.27.2 Weekly electrical bus surveillance 23 SO23-3-3.60 Inservice pump testing program 10 SO23-3-3.60.2 Low pressure safety injection pump 2(3)MP-016 Miniflow Testing * U3, P016, Quarterly Performance, 05/26/11, 04/23/10, (Rev. 9) * U3, P016, Quarterly Performance Trend Data, 06/01/08 - 02/26/11

• U3, P016, Comprehensive Test, 10/23/10, (Rev. 9) 9, 10 SO23-3-3.60.4 Saltwater cooling pump 2(3)mp-307 and v alve testing

• U2, P307, Quarterly Performance, 02/28/11, 03/30/11, 05/15/11 * U2, P307, Quarterly Performance Trend Data, 6/01/08 - 2/26/11

- 15 -

Attachment

Procedures

NAME TITLE REVISION / DATE

SO23-403-12-278 Commercial Instruction and Parts Manual 4700 kW Generator Set W.O. 73402, Ideal Electric,

DS-841 April 4, 1977 SO23-4-2-8 Saltwater injection temperature vs. Minimum saltwater flow, (operating instruction)

SO23-6-3 480 volt Switchgear Operation 14 SO23-I-2.90

DG Fuel Oil Storage Tank Water Accumulation Surveillance

SO23-II-11.110 Inverter & isolimiter inspection and cleaning 1 SO23-II-11.185 Test procedure for vital buses test and calibrartion for inverters Y001, Y002, Y003, and Y004

SO23-V-3.4 Inservice testing of pumps program 20 SO23-V-3.5.4 Inservice Testing of Check Valves 11 SO23-V-3.50 Administration of the Generic Letter 89-10 MOV Program 9

SO23-XXXIII-10 System engineering procedure 4

Surveillances NUMBER TITLE REVISION / DATE S023-3-3.30.1 RWST Outlet Valve Tests September 10, 2009 November 23, 2009 S023-3-3.60.1 Check Valve Partial Open Testing December 25, 2007 S023-I-6.20 TRW and Techno Corporation and Crane Nuclear Twin Flapper Check Valve Overhaul January 31, 2007 January 9, 2010 SO123-I-9.30 MOV Analysis and Test System (2HV9301) March 3, 2004

SO123-I-9.30 MOV Analysis and Test System (2HV9302) February 10, 2006 SO123-I-9.5 Electrical Inspection of Limitorque Actuators (2HV9302) February 10, 2006 SO23-3-3.23 Diesel Generator Monthly Surveillance July 5, 2011

- 16 -

Attachment Surveillances NUMBER TITLE REVISION / DATE SO23-3-3.30.1 Train B Safety Injection Valve Test August 9, 2009 April 17, 2010 July 12, 2010 October 3, 2010 December 20, 2010 March 24, 2011 June 12, 2011 SO23-3-3.31.3

CCW Train B Nitrogen Backup Regulator Test

December 9, 2009 SO23-3-3.31.6 Auxiliary Feedwater System Check Valve Test November 26, 2007 November 30, 2007 September 25, 2009 February 2, 2010 April 6, 2010

SO23-3-3.43.8 ESF Subgroup Relays K-109B and K-104B Semiannual Test February 26, 2011 SO23-3-31-10

Surveillance Test-Valve 10SWV-HV-6496

February 6, 2010 SO23-I-8-235 Cold Bench Testing of IST Safety-Relief Valves April 9, 2002

Specifications

NUMBER TITLE REVISION / DATE SO23-407-04 Specification for Vessels, Shop Fabricated October 1973

SO23-410-6-1-2 Seismic Qualification of Class I Ducting August 1975

Work Orders

800048917

800313086

800475366

800670622

800049321

800358730

800521887 03063187000

800049394

800369766

800607282 05032224000

800194370

800369797

800621634 06040160000

800235858

800438783

800635163 07030240000

800261784

800461765

800666862 32448313000

- 17 -

Attachment Miscellaneous

NUMBER TITLE REVISION / DATE

Guideline on component evaluation and classification to INPO

AP-913 REV. 4.0 January 26, 2009

ABB Power T&D Company Inc. Report of Transformer Tests S/O 24-34274S04 March 9, 1993

San Onofre Nuclear Generating Station, Unit 2 & 3 (UFSAR and UFHA - Amended June 2011)

San Onofre Nuclear Generating Station System Health Reports 1

st , 2 nd , 3 rd, and 4 th Quarter 2010

System Health Report, Emergency Diesel Generators Q4 - 2010 Q1 - 2011

System Health Report, Auxiliary Feedwater System Q4 - 2010

System Health Report, ECCS Q4 - 2010

Emergency Diesel Generator Recovery Plan 1

Instruction Manual for Wye Lift Check Valve March 20, 1991

IST Detail Report for 2HV9301 June 21, 2011

IST Detail Report for S21305MU124 June 21, 2011

IST Detail Report for 2HV9302 June 21, 2011

IST Detail Report for S21204MU001 June 21, 2011

Songs system health report - 120 V and lighting - voltage systems: 4th quarter 2010 No Date

Cable vault inspection spread sheet

March/April 2011

NCDB master equipment list system eroute raceway detail report for all non-safety related manholes containing inaccessible cables including EP and security cables July 6, 2011

Estimate of Tsunami Effect at SONGS December 31, 1972

070800412 Operating Experience Evaluation, NRC

IN 2007-27 August 8, 2007 1814-AE707-M0009 Instruction Manual, Ultrasonic Flowmeters September 2008, (R/1) 2010-125-C0001 Calibration of Panametrics

PT 878 Flometer (Test 595163) April, 2010 2PE-517-01 EDG Building HVAC Balance, Unit 2 October 8, 1981

- 18 -

Attachment Miscellaneous

NUMBER TITLE REVISION / DATE

Byron Jackson letter to Southern California Edison Unit 2&3 Salt Water Cooling Pumps February 1, 1991 Byron Jackson Pump Division Letter to Bechtel Test Report on Vibration & Minimum Submergence, Salt Water Cooling Pumps March 15, 1979 (original) CCN NO.

CCN-1 Calculation change notice for calculation M-0073-

2 November 18, 1998 CCN NO.

CCN-3 Calculation change notice for calculation J-MA-015 October 10, 2000 CCN NO.

CCN-5 Calculation change notice for calculation M-0073-

2 April 30, 2004 CCN NO. D0010762 Calculation change notice for calculation J-RNA-

015 February 26, 2009 CCN NO. D0037803 Calculation change notice for calculation J-RNA-

015 November 17, 2010 CCN NO. D0046798 Calculation change notice for calculation E4C-051.1 March 29, 2011 Component Classification Evaluation Document No.

70593 Unit 2(3) Safety Injection System Orifice Plates, Models:

OP-FTT-316-300 &

OP-316-1500 May 8, 1992 Controlotron Telefax TO Southern CA

Edison Response to SONGS Questions, in Calculation J-EPA-002, Ref. 6.12 September 27, 1991

CS-15400-2 Instruction Manual Multi Phase Electronic Temperature Monitor for Dry Type Transformers 108-050-011CS-15400

D-031 Equipment reliability improvement intolerance to failure policy

DBD-SO23-140 FIGURE D-2 Simplified one line diagram, class 1E 125 Vdc system, Unit 3

DBD-SO23-360 Reactor Coolant System 11

DBD-SO23-410 Saltwater cooling system design bases document 9

DBD-SO23-740 Design Bases Document, Safety Injection, Containment Spray, and Shutdown Cooling Systems 10

- 19 -

Attachment Miscellaneous

NUMBER TITLE REVISION / DATE

DBD-SO23-780 Design Bases Document, Auxiliary Feedwater System 9 DCN

NT-2 Non-technical design change notice to calculation J-RNA-015 November 17, 1999 DCP NO. 793.01J Delete High Temperature Lockout from the Open and Close Circuits of the RCS Pumps Seal CCW Return Valves. December,16/1982 ECN NO. A44267 Engineering change notice for calculation J-RNA-

015 June, 29 2006 ECN NO. A51015 Engineering change notice for calculation J-RNA-

015 August, 29 2007 ECN NO. A51420 Engineering change notice for calculation J-RNA-

015 October, 4 2007 ECN NO. A52027 Engineering change notice for calculation E4C-051.1 October 17, 2007 ECN NO. A52155 Engineering change notice for calculation J-RNA-

015 November 6, 2007 ECN NO. A52870 Engineering change notice for calculation J-RNA-

015 December 6, 2007 ECN NO. A54850 Engineering change notice for calculation J-RNA-

015 March 7, 2008 ECN NO. D0011126 Engineering change notice for calculation E4C-051.1 January, 28 2011 ECN NO. D0011952 Engineering change notice for calculation E4C-051.1 April 29, 2009 ECN NO. D0012041 Engineering change notice for calculation E4C-051.1 April 29, 2009 ECN NO. D0015144 Engineering change notice for calculation E4C-051.1 July 15, 2009 ECN NO. D0015162 Engineering change notice for calculation E4C-051.1 July 16, 2009 ECN NO. D0015896 Engineering change notice for calculation E4C-051.1 July 22, 2009 ECN NO. D0021568 Engineering change notice for calculation E4C-051.1 January 7, 2010 ECN NO. D0027872 Engineering change notice for calculation E4C-051.1

December 10, 2010

- 20 -

Attachment Miscellaneous

NUMBER TITLE REVISION / DATE

ECN NO. D0038764 Engineering change notice for calculation E4C-051.1

October 7, 2010 ECN NO. D0050075 Engineering change notice for calculation E4C-051.1 June 8, 2011

ECP 040201281-

Transformer Cooling Fans Kit at Load Center

3B06X 0 FCD LMAIM

1401 Limitorque SMB Series/SB Series Installation and Maintenance

I.J.5.N-SCN 80071683-0020 10CFR 50.59 Screen for NECPs

800071683 and NECP

800071684

IPE-SA-004 Containment spray and eccs analysis (SONGS living PRA)

IPE-SA-010 Low prsssure safety injection system analysis (SONGS living PRA)

IPE-SA-016 Saltwater cooling system analysis 2 M-U2EF01 Maintenance Rule Monthly Functional Failure Report (EDG) 2008 - 2011

NECP

800071683 Remove the low-low level trip for diesel generator fuel transfer pumps

Unit 2

NECP

800071684 Remove the low-low level trip for diesel generator fuel transfer pumps Unit 3

NMO-800052533 Replacement of Diesel Generator feeder cable tunnel manhole flood level detector switch 3l Sh 9487-2 May 2, 2011

NMO-800052535 Replacement of Diesel Generator feeder cable tunnel manhole flood level detector switch 3l Sh 9489-2 June 23, 2010

NMO-800082532 Replacement of Diesel Generator feeder cable tunnel manhole flood level detector switch 3l Sh 9486-2 June 18, 2010

NMO-800082619 Replacement of 2l Sh 9488-2 flood level float switch May 18, 2011

NPF-10/15-514 Increase in Reactor Power to 3438 MWt May 31, 2001 NRC Information Notice 89-54 Potential Overpressurization of the Component Cooling System June 23, 1989

- 21 -

Attachment Miscellaneous

NUMBER TITLE REVISION / DATE

NRC INFORMATION NOTICE 97-31 Failures of Reactor Coolant Pump Thermal Barriers and Check Valves in Foreign Plants June 3, 1997 NRME -800050987 Control building flood level detector switch calibration 2l Sh 9476-2 and 2l Sh 9477-2

April 25, 2009 NRME-

800051156 Dg feeder cable tunnel manhole flood level detector level calibration 3lsh9486-2, 3l Sh 9487-2, 3l Sh 9488-2, 3l Sh 9489-1, and 3l Sh 9490-1 July 25, 2009 NRME

800143029 Molded case breaker inspection

RCM-CRI February 20, 2010 NRME

800172147 Inspect cable vaults for water intrusion May 13, 2009 NRME

800311535 Inspect cable vaults for water intrusion August 3, 2010

NRME-800368810 120 Vac bus/switch clean/inspect-megger January 9, 2011

NRME-800655572 Perimeter public address system (PPAS) q July 15, 2011

NRME-800198521 3y002 inverter inspect and clean December 6, 2010

NRME-800369212 3y002 inverter test and calibration PM December 24, 2010

NRME-800480710 3y002 and 3y004 filter PM September 16, 2010

NRME-800727078 Rejected 3y002 inverter inspection and calibration June 22, 2011

NRME-800727079 3y002 and 3y004 filter PM June 22, 2011

NRME-800727280 3y002 inverter/vital bus test and calibration June 22, 2011

NRME-800727281 Inverter capacitor changeout ac and dc January 2, 2011

NRME-800727282 Inverter capacitor changeout ac and dc June 22, 2011 Report No. 10CFR50-0077 Graham-White Air Start Solenoid Valves #712-065 and #712-015

January 22, 1998

- 22 -

Attachment Miscellaneous

NUMBER TITLE REVISION / DATE

S023-200-A,

450-C

Bechtel Letter-Seismic Anchor Movement September 19, 1978 S023-955-20-4 Vendor Manual; 20 kVa Uninterruptable Power Supply June 29, 1979 S2-1204-ML-004 Letter, SCE to NRC, Risk-Informed Inservice Testing and

GL 96-05 (Supplemental Information) September 29. 1999 S-269 Joy Manufacturing Co.: Report of Fan Seismic Qualification October 1980

SCE-NRC-9-6-94 Salt Water Cooling Emergency Discharge Licensing Basis September 6, 1994

SD-S023-410 Saltwater cooling system, description 7

SD-SO23-120

6.9 kV, 4.16 kV

AND 480 V Electrical Distribution System 19

SD-SO23-120 6.9 kV, 4.16 kV

AND 480 V Electrical Distribution Systems 20

SD-SO23-130 120 Vac class 1E electrical distribution system 12

SD-SO23-360 Reactor Coolant System, Description 10

SD-SO23-400 Component Cooling Water System Description 20

SD-SO23-612 Flood protection and underground edg piping leak detection systems

SD-SO23-720 Engineered Safety Features Actuation System 8

SD-SO23-740 Safety Injection, Containment Spray, And Shutdown Cooling Systems

SD-SO23-750 Emergency Diesel Generators 19

SEE-990010 Unit 2/3 Emergency Diesel Generator Air Start Solenoid Valves Sequence

2834096 Eroute raceway detail report

SO123-V-1.30 Cable aging management 0 SO23-12-1 Standard Post Trip Actions 22 SO23-12-11 EOI Supporting Attachments 8 SO23-12-7 Loss Of Forced Circulation/Loss Of Offsite Power 21

- 23 -

Attachment Miscellaneous

NUMBER TITLE REVISION / DATE

SO23-13-17 Recovery From Inadvertent Safety Injection/Containment Isolation Or Containment Spray 6 SO23-13-7 Loss Of Component Cooling Water (CCW) /Saltwater Cooling (SWC)

SO23-1-5 Auxiliary Building Normal HVAC System Operation 22 SO23-15-56.C Annunciator Panel 56C, RCP/CPC 17 SO23-15-60.A1 Annunciator Panel 60A, Emergency HVAC 12

SO23-410-6-27 Joy manufacturing company fan performance data sheets for fan S21510 MA165 (9 sheets) March 4, 1977 SO23-6-6 Reserve Auxiliary Transformer Operation 15 SO23-933-61 DCN, Instructions For 8x20wfd LPSI Pumps And 8XWFD CS Pumps, (4 Hrs Operation) June 14, 1996 Songs Living 2(3)

PRA Probabilistic Risk Assessment (PRA) Executive Summary Report Living-Current as of

2011 SONGS Maintenance Order 01100777000 120 Vac bus/switch clean/inspect-megger March 19, 2003 SONGS Maintenance Order 03100097000 Control building electrical manhole flood detector level calibration 2l Sh 9477-2 April 25, 2006 SONGS Maintenance Order 03120517000 Dg feeder cable tunnel manhole flood detector level calibration 3l Sh 9486-2, 3l Sh 9487-2, 3l Sh 9488-

2, 3l Sh 9489-1, and 3l Sh 9490-1 May 11, 2006 SONGS Maintenance Order

05011304000 DG feeder cable tunnel manhole flood level detector level calibration 2l Sh 9485-2, 2l Sh 9487-2, 2l Sh 9490-1, 2l Sh 9491-1, and 2l Sh 9492-1 June 21, 2007 SONGS Maintenance Order 05101427000 120 Vac bus/switch clean/inspect-megger December 6, 2006 SONGS System Health Report

SWCS - Salt Water Cooling System 1

st Quarter - 2011

- 24 -

Attachment Miscellaneous

NUMBER TITLE REVISION / DATE

SONGS System Health Report CCWS -

Component Cooling Water System 4

th Quarter - 2010 VIRP V-C283-

001 C & S Valves - Dual Plate Check Valve Instruction Manual 0