ML12352A067

Diablo Canyon, Units 1 and 2, Issuance of Amendment Nos. 215 and 217, Revise Technical Specifications Related to Auxilary Feedwater System to Adopt TSTF-245, Rev. 1, TSTF-340, Rev. 3, TSTF-412, Rev. 3, and TSTF-439, Rev. 2 (TAC Nos. ME6360

ML12352A067
Person / Time
Site:	Diablo Canyon
Issue date:	01/31/2013
From:	Joseph Sebrosky Plant Licensing Branch IV
To:	Halpin E D Pacific Gas & Electric Co
Sebrosky J M
References
TAC ME6360, TAC ME6361
Download: ML12352A067 (38)

Text

UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 January 31, 2013 Mr. Edward D. Halpin Senior Vice President and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 56, Mail Code 104/6 Avila Beach, CA 93424 DIABLO CANYON POWER PLANT, UNIT NOS. 1 AND 2 -ISSUANCE OF AMENDMENTS RE: REVISION TO TECHNICAL SPECIFICATION (TS) 3.6.6, "CONTAINMENT SPRAY AND COOLING SYSTEMS," TS 3.7.5, "AUXILIARY FEEDWATER (AFW) SYSTEM," TS 3.B.1, "AC SOURCES -OPERATING," TS 3.B.9, "DISTRIBUTION SYSTEMS -OPERATING," AND TS EXAMPLE 1.3-3 (TAC NOS. ME6360 AND ME6361)

Dear Mr. Halpin:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 215 to Facility Operating License No. DPR-BO and Amendment No. 217 to Facility Operating License No. DPR-B2 for the Diablo Canyon Power Plant (DCPP), Unit Nos. 1 and 2, respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated June 1, 2011, as supplemented by letters dated February 6, May 31, August 6, and November 1,2012. The amendments revise TS 3.6.6, "Containment Spray and Cooling Systems," TS 3.7.5, "Auxiliary Feedwater (AFW) System," TS 3.B.1, "AC [Alternating Current] Sources -Operating," TS 3.B.9, "Distribution Systems -Operating," and TS Example 1.3-3 to clarify the operability of an AFW train during alternate alignments; establish conditions, required actions, and Completion Times when one steam supply to the turbine-driven AFW pump is inoperable concurrent with an inoperable motor-driven AFW train; and remove second Completion Times from the TSs. These changes are consistent with the guidance provided in Technical Specifications Task Force (TSTF) Travelers TSTF-245, Revision 1, "AFW train operable when in service," TSTF-340, Revision 3, "Allow 7 day Completion Time for a turbine-driven AFW pump inoperable," TSTF-412, Revision 3, "Provide Actions for One Steam Supply to Turbine Driven AFW/EFW [Emergency Feedwater] Pump Inoperable," and TSTF-439, Revision 2, "Eliminate Second Completion Times Limiting Time From Discovery of Failure to Meet an LCO [Limiting Condition for Operation]."

E. Halpin A copy of the related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice. Docket Nos. 50-275 and 50-323

Enclosures:

1. Amendment No. 215 to DPR-80 2. Amendment No. 217 to DPR-82 3. Safety Evaluation cc w/encls: Distribution via Listserv UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-275 DIABLO CANYON NUCLEAR POWER PLANT. UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No, 215 License No. DPR-80 The Nuclear Regulatory Commission (the Commission) has found that: The application for amendment by Pacific Gas and Electric Company (the licensee), dated June 1, 2011, as supplemented by letters dated February 6, May 31, August 6, and November 1, 2012, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comrl1ission's regulations and all applicable requirements have been satisfied. Enclosure 1

-2 Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. DPR-80 is hereby amended to read as follows: Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 215, are hereby incorporated in the license. Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions. This license amendment is effective as of its date of issuance and shall be implemented within 90 days. FOR THE NUCLEAR REGULATORY COMMISSION /-U*.f..Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License No. DPR-80 and Technical Specifications Date of Issuance: January 31, 2013 UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555*0001 PACIFIC GAS AND ELECTRIC DOCKET NO. DIABLO CANYON NUCLEAR POWER PLANT, UNIT AMENDMENT TO FACILITY OPERATING Amendment No. 217 License No. DPR-82 The Nuclear Regulatory Commission (the Commission) has found that: The application for amendment by Pacific Gas and Electric Company (the licensee), dated June 1, 2011, as supplemented by letters dated February 6, May 31, August 6, and November 1, 2012, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and The issuance ofthis amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. Enclosure 2

-2 Accordingly. the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment. and Paragraph 2.C.(2) of Facility Operating License No. DPR-82 is hereby amended to read as follows: Technical Specifications (SSER 32, Section 8)* and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B. as revised through Amendment No. 217. are hereby incorporated in the license. Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. except where otherwise stated in specific license conditions. This license amendment is effective as of its date of issuance and shall be implemented within 90 days. FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Facility Operating License No. DPR-82 and Technical Specifications Date of Issuance: January 31, Michael T. Markley. Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation 2013 ATTACHMENT TO LICENSE AMENDMENT NO. 215 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 217 TO FACILITY OPERATING LICENSE NO. DPR-82 DOCKET NOS. 50-275 AND 50-323 Replace the following pages of the Facility Operating License Nos. DPR-80 and DPR-82, and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Facility Operating License Nos. DPR-80 REMOVE INSERT Facility Operating License Nos. DPR-82 REMOVE INSERT Technical Specifications REMOVE INSERT 1.3-2 1.3-2 1.3-5 1.3-5 1.3-6 1.3-6 3.6-13 3.6-13 3.7-10 3.7-10 3.7-11 3.7-11 3.7-12 3.7-12 3.7-12a 3.8-1 3.8-1 3.8-2 3.8-2 3.8-29 3.8-29 Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility, This License shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: (1 Maximum Power Level The Pacific Gas and Electric Company is authorized to operate the facility at reactor core power levels not in excess of 3411 megawatts thermal (100% rated power) in accordance with the conditions specified herein, Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No, 215, are hereby incorporated in the license, Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions, Initial Test Program The Pacific Gas and Electric Company shall conduct the post-fuel-Ioading initial test program (set forth in Section 14 of Pacific Gas and Electric Company's Final Safety Analysis Report, as amended), without making any major modifications of this program unless modifications have been identified and have received prior NRC approval, Major modifications are defined as: Elimination of any test identified in Section 14 of PG&E's Final Safety Analysis Report as amended as being essential; Amendment No, 215 Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. This License shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: Maximum Power Level The Pacific Gas and Electric Company is authorized to operate the facility at reactor core power levels not in excess of 3411 megawatts thermal (100% rated power) in accordance with the conditions specified herein. Technical Specifications (SSER 32, Section 8)* and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 217, are hereby incorporated in the license. Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions. Initial Test Program (SSER 31! Section 4.4.1) Any changes to the Initial Test Program described in Section 14 of the FSAR made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change. *The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is . Amendment No. 217 1.3 Completion Times 1.3 Completion Times DESCRIPTION (continued) The total Completion Time allowed for completing a Required Action to address the subsequent inoperability shall be limited to the more restrictive of either: a. The stated Completion Time, as measured from the initial entry into the Condition, plus an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; or b. The stated Completion Time as measured from discovery of the subsequent inoperability. The above Completion Time extensions do not apply to those Specifications that have exceptions that allow completely separate entry into the Condition (for each train, subsystem, component, or variable expressed in the Condition) and separate tracking of Completion Times based on this re-entry. These exceptions are stated in individual Specifications. The above Completion Time extension does not apply to a Completion Time with a modified "time zero." This modified "time zero" may be expressed as a repetitive time (Le., "once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />," where the Completion Time is referenced from a previous completion of the Required Action versus the time of Condition entry) or as a time modified by the phrase "from discovery ..." (continued) DIABLO CANYON -UNITS 1 & 2 Unit 1 -Amendment No. 215 Unit 2 -Amendment No. 217 1.3 Completion Times 1.3 Completion Times EXAMPLES EXAMPLE 1.3-3 (continued) ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One Function X train inoperable. A.1 Restore Function X train to OPERABLE status. 7 days B. One Function Y train inoperable. B.1 Restore Function Y train to OPERABLE status. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> C. One Function X train inoperable. AND One Function C.1 Restore Function X train to OPERABLE status. OR 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Y train inoperable. C.2 Restore Function Y train to OPERABLE status. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ( continued) DIABLO CANYON -UNITS 1 & 2 1.3-5 Unit 1 -Amendment No. 436. 215 Unit 2 -Amendment No. 436. 217 1.3 Completion Times 1.3 Completion Times EXAMPLES EXAMPLE 1.3-3 (continued) When one Function X train and one Function Y train are inoperable, Condition A and Condition B are concurrently applicable. The Completion Times for Condition A and Condition B are tracked separately for each train starting from the time each train was declared inoperable and the Condition was entered. A separate Completion Time is established for Condition C and tracked from the time the second train was declared inoperable (Le., the time the situation described in Condition C was discovered). If Required Action C.2 is completed within the specified Completion Time, Conditions Band C are exited. If the Completion Time for Required Action A,1 has not expired, operation may continue in accordance with Condition A, It is possible to alternate between Conditions A, B, and C in such a manner that operation could continue indefinitely without ever restoring systems to meet the LCO. However, doing so would be inconsistent with the basis of the Completion Times. Therefore, there shall be administrative controls to limit the maximum time allowed for any combination of Conditions that result in a single contiguous occurrence of failing to meet the LCO. These administrative controls shall ensure that the Completion Times for those Conditions are not inappropriately extended. (continued) DIABLO CANYON -UNITS 1 & 2 1.3-6 Unit 1 -Amendment No. 215 Unit 2 -Amendment No. 217 -1 Containment Spray and Cooling Systems 3.6.6 3.6 CONTAINMENT SYSTEMS , 3.6.6 Containment Spray and Cooling Systems LCO 3.6.6 The containment fan cooling unit (CFCU) system and two containment spray trains shall be APPLICABILITY: MODES 1, 2, 3, and CONDITION REQUIRED ACTION COMPLETION TIME A. One containment spray train inoperable. A.1 OR A.2 Restore containment spray train to OPERABLE status. Restore containment spray train to OPERABLE status 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> For planned maintenance or inspections, the Completion Time is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The Completion Times of Required Action A.2 are for unplanned corrective maintenance or inspections. --.._--_.... --------------------..14 days B. Required Action and associated Completion Time of Condition A not B.1 AND Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> met. B.2 Be in MODE 5. 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> C. One required CFCU system inoperable such that a minimum of two CFCUs remain OPERABLE. C.1 Restore required CFCU system to OPERABLE status. 7 days (continued) DIABLO CANYON -UNITS 1 & 2 3.6-13 Unit 1 -Amendment No. m,m, 215 Unit 2 -Amendment No. m,4-'73,2@, 21l AFW System 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Auxiliary Feedwater (AFW) System LCO 3.7.5 Three AFW trains shall be OPERABLE. -----------------------------------------------NOT E Only one AFW train, which includes a motor driven pump, is required to be OPERABLE in MODE 4. MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal. ACTIONS --------------------------------------------N LCO 3.0.4b is not applicable. CONDITION REQUIRED ACTION COMPLETION TIME A. Turbine driven AFW train inoperable due to one inoperable steam supply. OR ------------NOT E Only applicable if MODE 2 has not been entered following refueling. Turbine driven AFW pump inoperable in IVIODE 3 following refueling. A.1 Restore affected equipment to OPERABLE status. 7 days B. One AFW train inoperable in MODE 1. 2 or 3 for reasons other than Condition A. B.1 Restore AFW train to OPERABLE status. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (continued) DIABLO CANYON -UNITS 1 & 2 3.7-10 Unit 1 -Amendment No. 215 Unit 2 -Amendment No. 217 AFWSystem 3.7.5 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. --------------NOTE Only applicable when the remaining OPERABLE motor driven AFW train provides feedwater to the steam generator with the C.1 Restore the steam supply to the turbine driven train to OPERABLE status. OR 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> inoperable steam supply. .._-_..._--_... ----------------_... _.....Turbine dr.iven AFW train inoperable due to one inoperable steam supply. AND One motor driven AFW train inoperable. C.2 Restore the motor driven AFWtrain to OPERABLE status. 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> D. Required Action and associated Completion Time of Condition A, B, or D.1 AND Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> C not met. OR Two AFW trains inoperable in MODE 1, 2 or 3 for reasons other than Condition C. D.2 Be in MODE 4. 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> E. Three AFW trains inoperable in MODE 1, 2, or 3. E.1 -----------N 0 TELCO 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status. ------.................................................... Initiate action to restore one AFW train to OPERABLE status Immediately F. Required AFW train inoperable in MODE 4. F.1 Initiate action to restore AFW train to OPERABLE status. Immediately DIABLO CANYON -UNITS 1 & 2 Unit 1 -Amendment No. 43e,4e9,215 Unit 2 -Amendment No. 43e,++G, 217 AFW System 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 -----------------------------NOTE AFW train(s) may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation. Verify each AFW manual, power operated, and automatic valve in each water flow path, and in both steam supply flow paths to the steam turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position. In accordance with the Surveillance Frequency Control Program SR 3.7.5.2 Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after 650 psig in the steam generator. Verify the developed head of each AFW pump at the flow test point is greater than or equal to the required developed head. In accordance with the Inservice Test Program. SR 3.7.5.3 -----------------------------NOTE AFW train(s) may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation. Verify each AFW automatic valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal. In accordance with the Surveillance Frequency Control Program (continued) DIABLO CANYON -UNITS 1 & 2 3.7-12 Unit 1 -Amendment No. 215 Unit 2 -Amendment No. 4J.a ,-t88,2-G+,2-00,217 AFW System 3.7.5 SURVEILLANCE REQUIREMENTS (continued) SR 3.7.5.4 ----------------------------N OTE Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after;::: 650 psig in the steam generator. AFW train(s) may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation. Verify each AFW pump starts automatically on an actual or simulated actuation signal. SR 3.7.5.5 Not used. In accordance with the Surveillance Frequency Control Pro ram DIABLO CANYON -UNITS 1 & 2 3.7-128 Unit 1 -Amendment No. 215 Unit 2 -Amendment No. 217 AC Sources -Operating 3.8.1 3.8 ELECTRICAL POWER SYSTEMS AC Sources -Operating LCO 3.8.1 The following AC electrical sources shall be OPERABLE: Two qualified circuits between the offsite transmission network and the onsite Class 1 E AC Electrical Power Distribution System; and Three diesel generators (DGs) capable of supplying the onsite Class 1 E power distribution subsystem( s}; and Two supply trains of the diesel fuel oil (DFO) transfer system. APPLICABILITY: MODES 1, 2, 3. and 4. ACTIONS LCO 3.0.4b is not applicable to DGs. CONDITION A. One required offslte circuit A.1 inoperable. ANDA.2 REQUIRED ACTION COMPLETION TIME Perform SR 3.8.1.1 for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> required OPERABLE ANDoffsite circuit. Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter. Restore required offsite 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> circuit to OPERABLE status. (continued) DIABLO CANYON -UNITS 1 &2 3.8-1 Unit 1 -Amendment No. m.m,400-, 215 Unit 2 -Amendment No. m . .:t-e+,-1+G, 217 -l AC Sources" Operating 3.8.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. One DG inoperable. B.1 AND B.2 AND B.3.1 B.3.2 AND IB.4 Perform SR 3.8.1.1 for the required offsite circuit(s). ------------N0 TE In MODE 1, 2, and 3, TDAFW pump is considered a required redundant feature . ... _-----_... __.... _-_.... Declare required feature( s) supported by the inoperable DG inoperable when its required redundant feature(s) is inoperable. Determine OPERABLE DG(s) is not inoperable due to common cause failure. OR Perform SR 3.8.1.2 for OPERABLE DG(s). Restore DG to OPERABLE status. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> AND Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from discovery of Condition B concurrent with inoperability of redundant required feature( s). 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 24 hours 14 days (continued) DIABLO CANYON -UNITS 1 & 2 3.8-2 Unit 1 -Amendment No. 215 Unit 2 -Amendment No. m,.:J-*i.+, 217 Distribution Systems -Operating 3.8.9 3.8 ELECTRICAL POWER SYSTEMS 3.8.9 Distribution Systems-Operating LCO The required Class 1E AC, DC, and 120 VAC vital bus electrical power distribution subsystems shall be OPERABLE. MODES 1, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One AC electrical power distribution subsystem inoperable. A.1 Restore AC electrical power distribution subsystem to OPERABLE status. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> B. One 120 VAC vital bus subsystem inoperable. B.1 Restore 120 VAC vital bus subsystem to OPERABLE status. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> C. One DC electrical power distribution subsystem inoperable. C.1 Restore DC electrical power distribution subsystem to OPERABLE status. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> D. Required Action and associated Completion Time not met. D.1 AND D.2 Be in MODE 3. Be in MODE 5. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours E. Two required Class 1 E AC, DC. or 120 VAC vital buses with inoperable distribution subsystems that result in a loss of safety function. E.1 Enter LCO 3.0.3. Immediately DIABLO CANYON -UNITS 1 & 2 3.8-29 Unit 1 -Amendment No. 43e, 215 Unit 2 -Amendment No..217 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 215 TO FACILITY OPERATING LICENSE NO. DPR-BO AND AMENDMENT NO. 217 TO FACILITY OPERATING LICENSE NO. DPR-B2 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON POWER PLANT. UNIT NOS. 1 AND 2 DOCKET NOS. 50-275 AND 50-323

1.0 INTRODUCTION

By application dated June 1, 2011 (Agencywide Documents Access and Management (ADAMS) Accession No. ML 111530114), as supplemented by letters dated February 6, May August 6, and November 1, 2012 (ADAMS Accession Nos. ML 1203BA041, ML ML 12220A164, and ML 12306A533, respectively), Pacific Gas and Electric Company the licensee) requested changes to the Technical Specifications (TSs) (Appendix A to Operating License Nos. DPR-BO and DPR-B2) for the Diablo Canyon Power Plant Unit Nos. 1 and The amendments would revise TS 3.6.6, "Containment Spray and Cooling Systems," TS "Auxiliary Feedwater (AFW) System," TS 3.B.1, "AC [Alternating Current] Sources -TS 3.B.9, "Distribution Systems -Operating," and TS Example 1.3-3 to clarify the operability of AFW train during alternate alignments; establish conditions, required actions, and Times when one steam supply to the turbine-driven AFW pump is inoperable concurrent with inoperable motor-driven AFW train; and remove second Completion Times from the TSs. changes are consistent with the guidance provided in Technical Specifications Task (TSTF) Travelers TSTF-245, Revision 1, "AFW train operable when in service," Revision 3, "Allow 7 day Completion Time for a turbine-driven AFW pump inoperable," Revision 3, "Provide Actions for One Steam Supply to Turbine Driven AFW/EFW Feedwater] Pump Inoperable," and TSTF-439, Revision 2, "Eliminate Second Completion Limiting Time From Discovery of Failure to Meet an LCO [Limiting Condition for The supplemental letters dated February 6, May 31, August 6, and November 1,2012, additional information that clarified the application, did not expand the scope of the as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) original proposed no significant hazards consideration determination as published in the Register on December 13, 2011 (76 FR Enclosure 3 2.0 REGULATORY The NRC staff considered the following regulatory requirements and guidance in its review of the licensee's application. The Commission's regulatory requirements related to the content of the TSs are set forth in Section 50.36, "Technical specifications," of Title 10 of the Code of Federal Regulations (10 CFR). Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. The regulations in 10 CFR 50, Appendix A, General Design Criterion (GDC) 34, "Residual heat removal," specify, in part, that the "system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded." In 10 CFR 50, Appendix A, GDC 44, "Cooling water," specifies, in part, that a system shall be provided "to transfer heat from structures, systems, and components (SSCs) important to safety, to an ultimate heat sink." In 10 CFR 50.63, "Loss of all alternating current power," requires all licensees to assess the capability of their plant to maintain adequate core cooling and appropriate containment integrity during a Station Blackout and to have procedures to cope with the event. In 10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," requires each licensee to monitor the performance or condition of SSCs against licensee-established goals to ensure that the SSCs are capable of fulfilling their intended functions. Such goals shall be established commensurate with safety, and where practical, take into account industry-wide operating experience. The NRC staff reviewed the proposed changes for compliance with 10 CFR 50.36. In general, licensees must fully describe and justify the desired TS changes. The NRC staff then makes a determination as to whether the proposed changes maintain adequate safety. Changes that result in relaxation (less restrictive condition) of current TS requirements require detai/ed justification. Licensees may revise the TSs provided that plant-specific review supports a finding of continued adequate safety because: (1) the change is editorial, administrative or provides clarification (Le., no requirements are materially altered), (2) the change is more restrictive than the licensee's current requirement, or (3) the change is less restrictive than the licensee's current requirement, but nonetheless still affords adequate assurance of safety when judged against current regulatory standards. The detailed application of this general framework, and additional specialized guidance, are discussed in Section 3.0 in the context of specific proposed changes.

-3

3.0 TECHNICAL EVALUATION

3.1 Description of Technical Specification Changes 3.1.1 ChangestoTS3.7.5 Current Condition A states: One steam supply to turbine driven AFW pump inoperable. Revised Condition A would state: Turbine driven AFW pump train inoperable due to one inoperable steam supply. OR --------------------------------------------NOT E -----------------------------------------Only applicable if MODE 2 has not been entered following refueling. Turbine Driven AFW pump inoperable in MODE 3 following refueling. Current Required Action A.1 states: Restore steam supply to OPERABLE status. Revised Required Action A.1 would state: Restore affected equipment to OPERABLE status. In addition, the licensee proposes to delete the second Completion Time from Conditions A and B, which currently state "10 days from discovery of failure to meet the LCO."

-New Condition C1 would state: ()nly applicable when the remaining ()PE:RABLE: motor driven AFW train provides feedwater to the steam generator with the inoperable steam supply. Turbine driven AFW train inoperable due to one inoperable steam supply. ()ne motor driven AFW train inoperable. New Required Actions for Condition C would state: C.1 Restore the steam supply to the turbine driven train to ()PE:RABLE: status. C.2 Restore the motor driven AFW train to ()PE:RABLE: status. As a result of adding new Condition C, current Conditions C through E: would be renumbered as Conditions D through F. In addition, current Condition C, renumbered as Condition D, states: Required Action and associated Completion Time for Condition A or B not met. Two AFW trains inoperable in M()DE: 1, 2 or 3. In the licensee's letter dated June 1, 2011, this proposed condition was originally identified as Condition D. By letter dated November 1, 2012, the licensee withdrew its request to add a new proposed TS 3.7.5 Condition B that was part of the June 1,2011, submittal. Because of the removal of proposed TS 3.7.5 Condition B, proposed new TS 3.7.5 Condition D was renumbered as new Condition C. Regarding proposed new TS 3.7.5 Condition B, in its June 1,2011, submittal, PG&E proposed this new Condition and associated TS bases section to address a failure of automatic level control affecting level control valves on motor-driven AFW trains; however, by letter dated November 1, 2012, PG&E removed this portion of the license amendment request. The removal of this portion of the June 1, 2011, request was based on a design change installed in DCPP, Unit 1, and scheduled to be installed in Unit 2 during Unit 2 Refueling Outage 17 that affects the input to the AFW automatic level control system. The design change renders the need for the TS change unnecessary. In the November 1,2012, submittal, PG&E proposed a new commitment to install the design change in Unit 2. This commitment is discussed in Section 4.0 of this safety evaluation.

-Revised Condition D would state: Required Action and associated Completion Time of Condition A, B, or C not met. Two AFW trains inoperable in MODE 1, 2 or 3 for reasons other Condition A new NOTE would be added to current SR 3.7.5.1 and would state: AFW train(s) may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually realigned to the AFW mode of operation. The above NOTE would also replace the SR 3.7.5.3 and SR 3.7.5.4 NOTES, which state, "Not applicable in MODE 4 when generator is relied upon for heat removal." Other TS Changes The licensee proposed to modify other TSs which includes the following: Revise TS example 1.3-3 to remove the second Completion Times from Required Actions A.1 and B.1 and to modify the discussion in that example (see Section 3.6 of this safety evaluation). Remove the second Completion Times associated with TS 3.6.6 Required Actions A.1, A.2, and C.1; TS 3.8.1 Required Actions A.2 and S.4; and TS 3.8.9 Required Actions A.1, B.1, and C.1. Change the TS 3.8.1 Completion Time for Condition B.4 from "14 days from discovery of failure to meet LCO" to "14 days." System Description for AFW In the Final Safety Analysis Report Update (FSARU) Chapter 6, Section 6.5, the licensee provides a description of the AFW system and its safety function. The licensee credits the AFW system to automatically supply sufficient water to the four steam generators (SGs) to remove decay heat upon the loss of normal feedwater. The AFW system has one 1 driven AFW (TDAFW) pump and two 50-percent motor-driven AFW (MDAFW) pumps. The two MDAFW pumps are powered from separate emergency buses. Each MDAFW pump is designed to deliver a minimum of 390 gallons per minute (gpm) to two of the four SGs. The TDAFW pump is designed to provide a minimum of 780 gpm divided between the four SGs. Steam from two of the four main steam lines provides the 'force for the TDAFW pump; however, steam from one main steam line is sufficient for the pump to obtain full flow.

The licensee states that provisions exist within the AFW system design to limit, control, or terminate AFW water flow to a faulted SG as necessary to: 1) prevent containment overpressurization following a steam line break inside containment 2) for steam leads 3 and 4, maintain the temperature profile in an area of the auxiliary building (area GE/GW) within analyzed limits, and 3) ensure minimum flow to the remaining unfaulted loops. Instrumentation on the MDAFW pumps' discharge lines, along with SG level input, feed into the Eagle 21 control system. If the system detected a depressurized SG, then it would signal the control valves to automatically throttle to limit flow to a depressurized SG to prevent a pump runout condition that may damage the MDAFW pump. The steam turbine pump discharge lines do not have this instrumentation; hence, manual operator action is credited for terminating AFW flow from a TDAFW pump to a depressurized SG. AFW is credited in the following Design Basis Events (DBEs): loss of main feedwater, secondary system pipe rupture, loss of all Alternating Current (AC) power, and loss-of-coolant accident (LOCA). In addition, the licensee credits AFW for the Reactor Coolant System (RCS) cooldown from a hot shutdown condition to the point Residual Heat Removal (RHR) system can be placed into service. In the FSARU, the licensee describes the conservative assumptions used in its accident analyses for the AFW system. The loss of main feedwater transient provides the basis for the minimum AFW flow requirement of 600 gpm to four SGs in order to prevent a release from the RCS pressurizer safety valve. The licensee assumes flow from both MDAFW pumps to initiate within 60 seconds after receiving a low-low SG level signal. For a rupture of main feedwater piping, the licensee states that AFW system design provides for terminating, limiting, or minimizing that fraction of AFW flow, which is delivered to a faulted loop or spilled through a break to ensure that sufficient flow is delivered to the remaining intact SGs. Initially, the licensee conservatively assumes maximum AFW flow to the ruptured SG at the time of the break, assuming pump runout protection fails. The licensee credits operator actions to isolate AFW flow to the ruptured SG within 10 minutes after the break, and to establish a minimum AFW flow of 390 gpm to two intact SGs. In the event of a main steam line rupture, initially the RCS rapidly cools down. For breaks inside containment, containment pressure and temperature increase. AFW is not initially required, but the licensee assumes conservatively maximum AFW flow continues to the faulted loop, adding to the mass and energy released to containment. The steam line rupture inside containment bounds the upper limit of AFW flow delivered to a faulted SG. After the initial SG blowdown, the RCS heats up again, at which time the licensee credits AFW flow to the intact SGs in order to limit the RCS heatup, but at lower flow rate than for the loss of feedwater transient. The licensee describes the worst-case break occurs in the steam supply piping to the TDAFW pump downstream of the non-return valve. A break at this location renders the TDAFW pump inoperable, and the steam from the rupture could potentially affect the operability of Level Control Valves (LCVs), LCV-113 and LCV-115, for the MDAFW pump number 3, rendering that pump inoperable. Then, a single failure of the remaining MDAFW pump number 2 could render the plant without AFW. Therefore, the licensee must either credit isolating the steam break or maintaining the level control valves LCV-113 and LCV-115 for MDAFW pump number 3 capable to operate in a steam harsh environment.

-7 In the event of a LOCA, AFW maintains the ability to provide water to the SGs to provide a heat sink for removing decay heat during RCS cooldown and depressurization to a cold shutdown condition. The AFW flow requirements are bounded by other accidents. In the event of a total loss of AC power, AFW flow is provided by the TDAFW pump to provide an adequate heat sink until AC power is restored. 3.3 Evaluation of TSTF-340. Revision 3 The licensee proposed changes to TS 3.7.5 in order to conform to the provisions in TSTF-340, Revision 3. The proposed change would allow a 7 -day Completion Time for the TDAFW pump being declared inoperable due to Surveillance Requirements (SRs) while in MODE 3, prior to entering MODE 2. The provision is only applicable if MODE 2 has not been entered following refueling. The extension to a 7 -day Completion Time would allow for additional time to repair and retest the TDAFW pump without cooling the RCS to MODE 4 to perform repairs and heating back to MODE 3 to retest the pump. The NRC staff approved Revision 3 to TSTF-340 in a letter dated March 16, 2000 (ADAMS Accession No. ML003694199). The NRC staffs technical justification for a longer Completion Time for an inoperable TDAFW pump in MODE 3 is acceptable based upon: 1) the minimal RCS decay heat levels upon entering MODE 3 after a refueling outage, 2) the redundant capabilities afforded by the MDAFW pumps, and 3) the low probability of a design basis accident requiring the operation of the TDAFW pump during the 7 -day Completion Time. The NRC staff has incorporated TSTF-340 into Revision 4 of NUREG-1431, "Standardized Technical Specifications -Westinghouse Plants." The AFW system at DCPP has three AFW pumps, two motor-driven and one turbine-driven AFW pumps. The 100-percent TDAFW pump is rated for 780 gpm, which can be divided between all four SGs. Each of the two 50-percent MDAFW pumps is capable of providing 390 gpm to only two SGs. Hence, in the event the TDAFW pump is unavailable, the AFW system maintains complete redundancy and independence with the remaining two MDAFW pumps to mitigate all the DBEs that would require AFW. In the most limiting event for AFW, a loss of main feedwater, the required flow of a minimum of 600 gpm to four SGs is met by one TDAFW pump or two MDAFW pumps. In accordance with 10 CFR 50.63, licensees are required to have the ability to mitigate a total loss of AC. In the event of a total loss of all AC, the licensee has established a coping time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to restore electrical power. In FSARU Section 6.5.2, the licensee credits the use of the TDAFW pump during those 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to maintain the plant at hot shutdown conditions. In the event the TDAFW pump is unavailable, the affected unit loses diversity of AFW pump's motive capabilities, leaving operators to rely solely on the remaining two electrically powered MDAFW pumps to mitigate events requiring AFW. However, the MDAFW pumps may not be available in a total loss of AC event. Currently, the TS 3.7.5 Required Action statement for an inoperable TDAFW pump at 100 percent power allows 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to restore the function. Hence, the NRC staff has evaluated the risk and consequences of having the TDAFW pump out of service at full power and found 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> acceptable. Allowing 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for one inoperable AFW pump is in agreement with the provisions in NUREG-1431, "Standard Technical Westinghouse Plants." The licensee is requesting to extend the allowed outage time from

-8 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days, when the plant is in MODE 3. During MODE 3, the consequences of having the TDAFW pump inoperable are less significant than at full power. The reactor has not been critical since shutdown for refueling; therefore, the RCS decay heat is at a minimum. Also, the SGs have an initial inventory of water that will remove decay heat by steaming through the main steam safety valves, allowing time for the operators to take remedial actions to restore power or take additional measures to restore the heat removal function. Specifically, at DCPP the other unit on-site is powered from a diverse offsite electrical supply and has its own three emergency diesel generators. Hence, there is a greater probability of restoring electrical power to the accident unit from the other unit's power sources in a short period of time prior to the fuel overheating. The NRC staff accepted the provisions of TSTF-340 allowing a 7-day Completion Time based on availability of the remaining MDAFW pumps and the reduced amount of decay heat following refueling prior to the reactor being critical. The NRC staff concludes that there is reasonable assurance that DCPP can successfully mitigate design-basis accidents while the TDAFW pump is unavailable, not allowing for an additional single failure. Based on the above, the NRC staff concludes that the adoption of TSTF-340, Revision 3, is acceptable for DCPP, Units 1 and 2. 3.4 Evaluation of TSTF-245. Revision 1 The licensee proposed changes to the action requirements stated in TS SR 3.7.5.1, SR 3.7.5.3, and SR 3.7.5.4 in order to conform to the provisions in TSTF-245, Revision 1. The proposed changes would add a note to the SRs that would allow an AFW train to be considered operable at low-power operation when its components are being operated manually for SG level control in MODES 2, 3, 4, and 5, and the components can be realigned for AFW mode of operation. The NRC staff approved TSTF-245, Revision 1, on July 3, 2003 (ADAMS Accession No. ML040611 028). The TSTF evaluation includes a letter from the NRC staff to Indian Point, dated May 23, 1997, which outlines the NRC staff's position on taking credit for manual actions. The letter states in part, "In general, it is not appropriate to take credit for manual action in place of automatic action for protection of safety limits to consider equipment operable." Thus, credit for any manual actions should be part of the plant's licensing basis. In order to credit manual actions, the licensee must evaluate physical differences between automatic and manual actions, and the ability to perform the manual actions. The letter states that the NRC staff has made a determination that for the AFW system on a typical pressurized-water reactor (PWR), manual actions versus automatic operation are permissible in certain circumstances. In particular, an AFW train may be considered fully operable when an operator is controlling AFW manually to maintain SG levels in the normal control band during startup, normal shutdown, and hot standby conditions. The NRC staff has incorporated the findings of TSTF-245, Revision 1, into TS SR 3.7.5.1 in Revision 4 of NUREG-1431, "Standard Technical Specifications -Westinghouse Plants." TS SR 3.7.5.1 requires that the licensee verify each AFW manual, power-operated, and automatic valve in each water flow path and in both steam supply flow paths to the TDAFW pump, that is not locked sealed or otherwise secured in position, is in the correct position. While the licensee is utilizing AFW in manual during low-power operations (Le., startups and shutdowns), the plant operators must take manual control of the AFW pump(s) and control

-9 valves to maintain proper SG level. In doing so, the valves may no longer be in the position assumed in the accident analyses. In FSARU Chapter 15 Accident Analyses, in the event of a feedwater line break, the licensee's analysis does not credit any AFW flow until AFW is isolated to the faulted SG at 10 minutes after the trip. Then the licensee credits AFW flow of a minimum feed rate of 390 gpm to two intact SGs. Therefore, the licensee's accident analysis provides a margin of 10 minutes for operator actions to align AFW to feed the intact SGs, or to isolate AFW flow to a faulted SG. For a main steam line break, the licensee's analysis assumes the worst initial conditions, where all AFW pumps are operating at full capacity, delivering 1700 gpm to the faulted SG and assumes AFW continues for the duration of the transient. These conditions result in the maximum RCS cooldown. Hence, there are no operator actions credited for isolating AFW flow. During plant startup, shutdown, and hot standby, operators are normally manually controlling AFW flow to the SGs. These operator actions are part of the normal operating procedures and the operators are trained to respond to plant transients. During these times, the AFW components would be temporarily incapable of automatic operations assumed in the accident analyses. Operators would have a margin of 10 minutes to correctly align AFW trains to meet the assumptions in the FSARU accident analyses. Ten minutes to manually isolate AFW to a faulted SG and manually align AFW to supply intact SGs during a feedwater line break is a reasonable amount of time for operators when operating AFW components in manual mode during startup, shutdown, and hot standby. Based on the above, the NRC staff concludes that the licensee's proposed changes are acceptable for the implementation of TSTF-245, Revision 1, for DCPP, Units 1 and 2. The AFW trains may be considered operable when operators have the AFW train's components in manual for SG level control during startup, shutdown, and hot standby, if the AFW components are capable of being manually realigned to the AFW mode of operation. 3.5 Evaluation of TSTF-412, Revision 3 The licensee proposes changes to TS 3.7.5 in order to incorporate similar provisions outlined in TSTF-412, Revision 3. The proposed change adds a Condition "C" that would create an action statement for one MDAFW pump to be inoperable coincident with one steam supply inoperable to the TDAFW pump for a period of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, if the remaining operable MDAFW train provides feedwater to the SG with the inoperable steam supply. TSTF-412 was approved as part of the Consolidated Line Item Improvement Process. The notice of availability for TSTF-412, Revision 3 was published in the Federal Register on June 17, 2007 (72 FR 39089). The NRC staff has incorporated TSTF-412 into Revision 4 of NUREG-1431, "Standard Technical Specifications -Westinghouse Plants." The proposed condition evaluates continued operations with one of the two steam supplies to the TDAFW pump inoperable and one MDAFW pump inoperable. The proposed 48-hour Completion Time for this condition is more restrictive than the Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for having one AFW pump inoperable, but it is less restrictive than having two AFW pumps inoperable, which requires the plant shutdown. In the TSTF evaluation, the NRC staff provides guidance on a

-Completion Time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> based upon the remaining operable MDAFW pump being able to supply 100 percent of the design flow as assumed in the accident analyses. The TDAFW pump can normally supply feedwater to all four SGs. It is supplied by two 1capacity steam lines from SG-2 and SG-3. Both steam supply lines are required for the TDAFW pump to be considered fully operable. With one steam supply initially inoperable, the TDAFW pump may not be able to perform its safety function if an accident occurs on the SG that supplies the one remaining steam line to the TDAFW pump. Each MDAFW pump can supply feedwater to only two SGs. The MDAFW pump 2 feeds SG-1 and SG-2; whereas, MDAFW pump 3 feeds SG-3 and SG-4. This uniqueness creates scenarios where one inoperable MDAFW pump and one inoperable steam supply to the TDAFW pump train, can lead to less than 100 percent AFW flow available to the SGs. For example, if the steam supply from SG-2 is inoperable to the TDAFW pump, and the MDAFW pump (2) that feeds SG-1 and SG-2 is inoperable at the same time, then a fault on SG-3 due to a feed line break or main steam line break renders the remaining steam supply to the TDAFW pump train unavailable; hence, it will not be able to perform its function. With SG-3 faulted, the remaining MDAFW pump 3 will only be able to feed SG-4. In this instance, the plant cannot meet it design basis requirement to deliver AFW flow to two SGs. By letter dated May 31,2012, the licensee responded to the NRC staff request for additional information on justifying a TS condition permitting a situation that may result in an unanalyzed condition under certain accident scenarios. The licensee proposes to restrict entering action statement "C" to "Only applicable when the remaining operable motor driven AFW train provides feedwater to the SG with the inoperable steam supply." In such configurations, in the event of a feed line break, the AFW system will continue to meet safety analysis flow requirements as described in the licensee's design basis (e.g., FSARU Table 6.5-2) which specifies in the event of a main feed line break, AFW flow to two of the four SGs is required. The licensee proposes a revised Condition "C" as follows: C. Only applicable when the remaining OPERABLE motor driven AFW train provides feedwater to the steam generator with the inoperable steam supply. C.1 Restore the steam supply to the turbine driven train to OPERABLE status. OR 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Turbine driven AFW train inoperable due to one inoperable steam supply. AND One motor driven AFW train inoperable. C.2 Restore the motor driven AFW train to OPERABLE status. 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> The noted restriction for entry would prevent the licensee from entering into unanalyzed conditions. The Completion Time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> affords sufficient time to restore an MDAFW pump or a steam supply to the TDAFW pump back to operable status. While in this degraded condition, the AFW system remains capable of mitigating the consequences of all design-basis

-11 accidents. Based on the above, the NRC staff concludes that the licensee's proposed Condition "C" is acceptable with the noted restriction on limited configurations for entry. 3.6 Evaluation of TSTF-439, Revision 2 On June 20,2005 (ADAMS Accession No. ML051860296), the commercial nuclear electrical power generation industry owners group, TSTF, submitted a proposed change, TSTF-439, Revision 2, "Eliminate Second Completion Times Limiting Time From Discovery of Failure To Meet an LCO," on behalf of the industry (TSTF-439, Revisions 0 and 1, were prior draft iterations). TSTF-439 proposed a new TS convention to limit the maximum time allowed for any combination of LCO Conditions that could result in a single continuous failure to meet the LCO. By letter dated January 11,2006 (ADAMS Accession No. ML060120272), the NRC approved TSTF-439, Revision 2. This traveler has been incorporated into Revision 4 of NUREG-1431, "Standard Technical Specifications -Westinghouse Plants." Second Completion Times are those joined by an "AND" logical connector to the specific Completion Time and which state, "X hours/days from discovery of failure to meet the LCO." At the time the format and usage rules of the Standard TS (STS) were created, there were no regulatory requirements or programs to preclude entry into and out of the actions for an indefinite period of time without meeting the LCO. A second Completion Time was included in the STS for certain Required Actions to establish a limit on the maximum time allowed for any combination of conditions that result in a single continuous failure to meet the LCO. The intent of the second Completion Time was to preclude entry into and out of the actions for an indefinite period of time without meeting the LCO. The second Completion Time provides a limit on the amount of time that an LCO would not be met for various combinations of conditions. The licensee's TSs are modeled after the STS. Section 1.3 of the licensee's TS, Example 1.3-3, describes the use of second Completion Times. The licensee proposed revising TS Example 1.3-3 to eliminate the second Completion Times in Conditions A and B, and to replace the discussion regarding second Completion Times with the following: It is possible to alternate between Conditions A, B, and C in such a manner that operation could continue indefinitely without ever restoring systems to meet the LCO. However, doing so would be inconsistent with the basis of the Completion Times. Therefore, there shall be administrative controls to limit the maximum time allowed for any combination of Conditions that result in a single contiguous occurrence of failing to meet the LCO. These administrative controls shall ensure that the Completion Times for those Conditions are not inappropriately extended. The licensee proposed deleting the following from the TS Example 1.3-3 description: Example 1.3-3 illustrates one use of this type of Completion Time. The 10 day Completion Time specified for Conditions A and B in Example 1.3-3 may not be extended.

-12 The licensee also proposed deleting the second Completion Times associated with TS 3.6.6 Required Actions A.1, A.2, and C.1, TS 3.7.5 Required Actions A.1 and B.1, TS 3.B:1 Required Actions A.2 and B.4, and TS 3.B.9 Required Actions A.1, B.1, and C.1. The licensee justified the proposed changes in its amendment request by referring to the controls of the Maintenance Rule, the Reactor Oversight Process, and the new requirement proposed in Section 1.3 of the TS. Specifically, the licensee stated: The Maintenance Rule: 10 CFR 50.65 (a)(1), the Maintenance Rule, requires each licensee to monitor the performance or condition of structures, systems, and components (SSCs) against licensee-established goals to ensure that the SSCs are capable of fulfilling their intended functions. If the performance or condition of an SSC does not meet established goals, appropriate corrective action is required to be taken. The NRC Resident Inspectors monitor the licensee's Corrective Action process and could take action if the licensee's maintenance program allowed the systems required by a single LCO to become concurrently inoperable multiple times. The performance and condition monitoring activities required by 10 CFR 50.65 (a)(1) and (a)(2) would identify if poor maintenance practices resulted in multiple entries into the ACTIONS of the TSs and unacceptable unavailability of these SSCs. The effectiveness of these performance monitoring activities, and associated corrective actions, is evaluated at least every refueling cycle, not to exceed 24 months per 10 CFR 50.65 (a)(3). Prior to 10 CFR 50.65, TSs were the primary rules governing operations, including what equipment must normally be in service, how long equipment can be out of service, compensatory actions, and surveillance testing to demonstrate equipment readiness. A goal of the TSs is to provide adequate assurance of the availability and reliability of equipment needed to prevent, and, if necessary, mitigate, accidents and transients. The Maintenance Rule shares this goal and operates through a dynamic and comprehensive process. The regulations in 10 CFR 50.65 assess and manage inoperable equipment; however, the rule also considers all inoperable risk-significant equipment. Under the TSs, the Completion Time for one system within an LCO is not generally affected by inoperable equipment in another LCO. However, the second Completion Time influenced the Completion Time for one system based on the condition of another system, but only if the two systems were required by the same LCO. Plant Maintenance Rule programs implement risk-based configuration management programs that augment the deterministic Completion Times in the TSs. The NRC Resident Inspectors also monitor the licensee's Corrective Action process and could take action if the licensee's maintenance program allowed the systems required by a single LCO to become concurrently inoperable multiple times. The performance and condition monitoring activities required by 10 CFR 50.65 identify maintenance practices that would result from multiple entries into the Actions of the TSs which contribute to unacceptable unavailability of these SSCs. The licensee's application states the following regarding the Reactor Oversight Process:

-NEI 99-02, "Regulatory Assessment Performance Indicator Guideline," describes the tracking and reporting of performance indicators to support the NRC's Reactor Oversight Process (ROP). The NEI document is endorsed by Regulatory Information Summary (RIS) 2001-11, "Voluntary Submission Of Performance Indicator Data." NEI 99-02, Section 2.2 describes the Mitigating Systems Cornerstone. NEI 99-02 specifically addresses emergency AC Sources (which encompasses the AC Sources Distribution System LCOs), and the Auxiliary Feedwater system. Extended unavailability due to multiple entries into the ACTIONS would affect the NRC's evaluation of the licensee's performance under the ROP. The objective of the cornerstone is to monitor the availability, reliability, and capability of systems that mitigate the effects of initiating events to prevent core damage. Licenses also reduce the likelihood of reactor accidents by maintaining the availability and reliability of mitigating systems. Mitigating systems include those systems associated with safety injection, decay heat removal, and their support systems, such as emergency AC power systems (which encompasses the AC Sources Distribution System LCOs, as noted by the licensee), and the AFW system. Inputs to the mitigating systems cornerstone include both inspection procedures and performance indicators to ensure that all ROP objectives are being met. Satisfactory licensee performance within the mitigation systems ROP cornerstone provides reasonable assurance in monitoring the inappropriate use of TS condition Completion Times. NRC inspection findings for each plant are documented in inspection reports in accordance with Inspection Manual Chapter (IMC) 0612, "Power Reactor Inspection Reports," and summarized in Plant Issues Matrices. Inspection findings are evaluated using the significance determination process in accordance with IMC 0609, "Significance Determination Process," to evaluate the safety significance of the findings. By letter dated February 6, 2012, the licensee stated the following regarding the requirement added in TS Section 1.3: ... , a requirement is added to Section 1.3 of the TSs to require licensees to have administrative controls to limit the maximum time allowed for any combination of Conditions that result in a single contiguous occurrence of failing to meet the LCO. These administrative controls should consider plant risk and shall limit the maximum contiguous time of failing to meet the LCO. PG&E will revise procedure OP1.DC17, "Control of Equip Required by Technical Specifications or Designated Programs." The administrative controls will ensure that a single contiguous occurrence of failing to meet the LCO will not be extended beyond the additive Completion Times of the two Required Actions for restoration unless a risk evaluation is performed, and the risk impact is managed. This TS requirement, when considered with the regulatory processes discussed above, provide an equivalent or superior level of plant safety without the unnecessary complication of the TSs by second Completion Times on some Specifications.

-The NRC staff reviewed the proposed changes and determined that they are less restrictive changes. However, NRC staff finds that multiple, continuous entries into TS conditions, without meeting the LCO, will be adequately controlled by the licensee's administrative controls and programs which were implemented to meet the requirements of the Maintenance Rule and is, therefore, acceptable. Specific to AFW, in Enclosure 2 of its letter dated June 1, 2011, the licensee provides an "example 4," which shows how the individual Completion Times would be limited to less than the 10-day second LCO time limit. These action statements were based upon the existing TS. The licensee did not assess incorporation of the provisions of TSTF-439 in conjunction with the incorporation of TSTF-412 and TSTF-340. The NRC staff requested the licensee to assess TSTF-439 with the conditions in TSTF-412 and TSTF-340. In a letter dated February 6,2012, the licensee's assessment of TSTF-439 under the new conditions determined a 7 -day period of inoperability will not be exceeded. The licensee's TS Section "Use and Application" requires the plant to remain in the first maximum allowed outage time constraint even if a secondary unrelated problem with the main component was discovered during this time period covered by a different allowed Completion Time. In TS Section 1.3, "Completion Times," the licensee states in the Description section: Once a Condition has been entered, subsequent trains, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition, unless specifically stated. The Required Actions of the Condition continue to apply to each additional failure, with Completion Times based on initial entry into the Condition. The licensee's modification (insert 6) to TS 1.3, "Completion Times," states, in part, that "Therefore, there shall be administrative controls to limit the maximum time allowed for any combination of Conditions that result in a single contiguous occurrence of failing to meet the LCO." For TS 3.7.5, Condition A, the Completion Time for both the inoperable TDAFW train due to one inoperable steam supply and the one inoperable TDAFW pump in MODE 3 following refueling is 7 days. The licensee must completely restore the pump to operable status within the original 7 -day Completion Time. Therefore, while in MODE 3, the licensee cannot extend the original 7 -day Completion Time by entering two separate 7 -day Completion Times for the different provisions of Condition A. The licensee must fully comply with the Required Actions within the initial entry's Completion Time or the licensee may extend the allowed Completion Time in accordance with the following guidance provided in Section 1.3 of the TS:

However, when a subsequent train, subsystem, component, or variable expressed in the Condition is discovered to be inoperable or not within limits, the Completion Time(s) may be extended. To apply this Completion Time extension, two criteria must first be met. The subsequent inoperability: a. Must exist concurrent with the first inoperability; and b. Must remain inoperable or not within limits after the first inoperability is resolved. The total Completion Time allowed for completing a Required Action to address the subsequent inoperability shall be limited to the more restrictive of either: The stated Completion Time, as measured from the initial entry into the Condition, plus an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; or The stated Completion Time as measured from discovery of the subsequent inoperability. Therefore, the licensee may be allowed to extend the 7 -day Completion Time an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, but the total time the licensee does not meet the lCO will not exceed the previous 10-day limitation, which was the intent of having a second Completion Time. The NRC staff concludes that the licensee's proposed changes satisfy applicable regulatory requirements, as described above in Section 2.0. In addition, the NRC staff concludes that assessment of the licensee's performance within the mitigation systems ROP cornerstone provides reasonable assurance in monitoring the inappropriate use of TS condition Completion Times. While the proposed changes are less restrictive than the licensee's current requirement, they still afford adequate assurance of safety when judged against current regulatory standards. Therefore, the NRC staff concludes that the licensee's proposed changes are acceptable for the implementation of TSTF-439, Revision 2, for DCPP, Units 1 and 2. REGULATORY COMMITMENTS In its letter dated June 1, 2011, and as revised and supplemented by its letter dated November 1,2012, the licensee made the following regulatory commitments: Commitment 1 PG&E will revise procedure OP1.DC17, "Control of Equip Required by Technical Specifications or Designated Programs." The administrative controls will ensure that a single contiguous occurrence of failing to meet the lCO will not be extended beyond the additive Completion Times of the two Required Actions for restoration unless a risk evaluation is performed, and the risk impact is managed. This TS requirement, when considered with the regulatory processes discussed above, provide an equivalent

-or superior level of plant safety without the unnecessary complication of the TSs by second Completion Times on some Specifications. Commitment 2 PG&E plans to install a control system median signal selector on DCPP Unit 2 during Unit 2 Refueling Outage Seventeen, currently scheduled to begin in February 2013, to address the design vulnerability where an Eagle-21 rack up can affect control of one level control valve per motor drive AFW train. The NRC staff concludes that reasonable controls for the implementation and for subsequent evaluation of proposed changes pertaining to the above regulatory commitments are best provided by the licensee's administrative processes, including its commitment management program. The above regulatory commitments do not warrant the creation of regulatory requirements (items requiring prior NRC approval of subsequent changes).

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the California State official was notified of the proposed issuance of the amendments. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration and there has been no public comment on such finding published in the Federal Register on December 13, 2011 (76 FR 77569). Accordingly, the amendments meet the eligibility criteria for categorical exclusion setforth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

-17

7.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public. Principal Stanley Gardocki Matthew Hamm Date: January 31, 2013 E. Halpin -A copy of the related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice. NRR/LPL4/DORULA NRR/DSS/SBPB/BCNRRlLPL4/DORUPM NRR/DSS/STSB/BC JBurkhardtJSebrosky GCasto RElliott* 12121/12 1114113 12/6/121/9/13I!E NRR/LPL4/DORUBC NRR/LPL4/DORLIPMOGC NLOFFICE MMarkley JSebroskyLSubinNAME 1/31/13 1/31/131/17/13DATE Docket Nos. 50-275 and 50-323

Enclosures:

1. Amendment No. 215 to DPR-80 2. Amendment No. 217 to DPR-82 3. Safety Evaluation cc w/encls: Distribution via Listserv DISTRI BUTION: PUBLIC LPLIV Reading RidsAcrsAcnw _MaiICTR Resource RidsNrrDorlDpr Resource RidsNrrDorlLpl4 Resource RidsNrrDssSbpb Resource RidsNrrDssStsb Resource RidsNrrLAJBurkhardt Resource RidsNrrPMDiabloCanyon Resource RidsOgcRp Resource RidsRgn4MailCenter Resource MHamm, NRRlDSS/STSB SGardocki, NRR/DSS/SBPB ADAMS Accession No. ML 12352A067 Sincerely, IRA! Joseph M. Sebrosky, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation *SE dated OFFICIAL RECORD