Text

Issuance of Amendment Nos. 236 and 238 Revision to Technical Specification 3.7.5, Auxiliary Feedwater (AFW) System, Exigent Circumstances
	ML20235R635
Person / Time
Site:	Diablo Canyon
Issue date:	08/31/2020
From:	Samson Lee; Plant Licensing Branch IV
To:	Welsch J; Pacific Gas & Electric Co
	Lee S, 301-415-3168
References
	EPID L-2020-LLA-0176
	Download: ML20235R635 (39)
	v • d • e

August 31, 2020 Mr. James M. Welsch Senior Vice President, Generation and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Nuclear Power Plant P.O. Box 56, Mail Code 104/6 Avila Beach, CA 93424

SUBJECT:

DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 236 AND 238 RE: REVISION TO TECHNICAL SPECIFICATION 3.7.5, AUXILIARY FEEDWATER (AFW) SYSTEM (EXIGENT CIRCUMSTANCES) (EPID L-2020-LLA-0176)

Dear Mr. Welsch:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 236 to Facility Operating License No. DPR-80 and Amendment No. 238 to Facility Operating License No. DPR-82 for the Diablo Canyon Nuclear Power Plant, Units 1 and 2, respectively. The amendments consist of changes to the technical specifications (TSs) in response to your application dated August 12, 2020, as supplemented by letters dated August 16, 2020; August 18, 2020; and August 20, 2020.

The amendments provide a new TS 3.7.5, Auxiliary Feedwater (AFW) System, action and continued operation of Unit 1 during Cycle 22 with the AFW system aligned in a manner for which current TS 3.7.5 would require a shutdown.

A copy of the related Safety Evaluation is enclosed. Notice of Issuance will be included in the Commissions biweekly Federal Register notice.

Sincerely,

/RA/

Samson S. Lee, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-275 and 50-323

Enclosures:

1. Amendment No. 236 to DPR-80

2. Amendment No. 238 to DPR-82

3. Safety Evaluation cc: Listserv

PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-275 DIABLO CANYON NUCLEAR POWER PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 236 License No. DPR-80

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Pacific Gas and Electric Company (the licensee), dated August 12, 2020, as supplemented by letters dated August 16, 2020; August 18, 2020; and August 20, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

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 Commissions regulations; D.

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 E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

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:

(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 236 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.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Jennifer L. Dixon-Herrity, 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: August 31, 2020 Jennifer L.

Dixon-Herrity Digitally signed by Jennifer L. Dixon-Herrity Date: 2020.08.31 08:56:05 -04'00'

PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-323 DIABLO CANYON NUCLEAR POWER PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 238 License No. DPR-82

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Pacific Gas and Electric Company (the licensee), dated August 12, 2020, as supplemented by letters dated August 16, 2020; August 18, 2020; and August 20, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

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 Commissions regulations; D.

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 E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

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:

(2)

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. 238, 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.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License No. DPR-82 and Technical Specifications Date of Issuance: August 31, 2020 Jennifer L.

Dixon-Herrity Digitally signed by Jennifer L. Dixon-Herrity Date: 2020.08.31 08:56:46 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 236 TO FACILITY OPERATING LICENSE NO. DPR-80 AND LICENSE AMENDMENT NO. 238 TO FACILITY OPERATING LICENSE NO. DPR-82 DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2 DOCKET NOS. 50-275 AND 50-323 Replace the following pages of Facility Operating License Nos. DPR-80 and DPR-82 and the 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 No. DPR-80 REMOVE INSERT Facility Operating License No. DPR-82 REMOVE INSERT Technical Specifications REMOVE INSERT 3.7-10 3.7-10 3.7-11 3.7-11 3.7-11a

Amendment No. 236 (4)

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 (5)

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.

C.

This License shall be deemed to contain and is subject to the conditions specified in the Commissions 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.

(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 236 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.

(3)

Initial Test Program The Pacific Gas and Electric Company shall conduct the post-fuel-loading 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:

a.

Elimination of any test identified in Section 14 of PG&Es Final Safety Analysis Report as amended as being essential;

Amendment No. 238 (4)

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 (5)

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.

C.

This License shall be deemed to contain and is subject to the conditions specified in the Commissions 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.

(2)

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. 238, 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.

(3)

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 discussed.

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.

NOTE---------------------------------------------------

Only one AFW train, which includes a motor driven pump, is required to be OPERABLE in MODE 4.

APPLICABILITY:

MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal.

ACTIONS

NOTE------------------------------------------------------

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

NOTE---------------

Only applicable if MODE 2 has not been entered following refueling.

Turbine driven AFW pump inoperable in MODE 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 or G.

B.1 Restore AFW train to OPERABLE status.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (continued)

DIABLO CANYON - UNITS 1 & 2 3.7-10 Unit 1 - Amendment No. 135,169, 215, Unit 2 - Amendment No. 135,170, 217, 236 238

AFW System 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 inoperable steam supply.

Turbine driven AFW train inoperable due to one inoperable steam supply.

AND One motor driven AFW train inoperable.

C.1 Restore the steam supply to the turbine driven train to OPERABLE status.

OR C.2 Restore the motor driven AFW train to OPERABLE status.

48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 48 hours D.

Required Action and associated Completion Time for Condition A, B, C, or G not met.

OR D.1 Be in MODE 3.

AND D.2 Be in MODE 4.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 18 hours Two AFW trains inoperable in MODE 1, 2 or 3 for reasons other than Condition C or G.

E.

Three AFW trains inoperable in MODE 1, 2, or 3.

E.1

NOTE---------------

LCO 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 (continued)

DIABLO CANYON - UNITS 1 & 2 3.7-11 Unit 1 - Amendment No. 135,169, 215, Unit 2 - Amendment No. 135,170, 217, 236 238

AFW System 3.7.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME G.

NOTE---------------

This Condition is only applicable to Unit 1 once during Unit 1 Cycle 22 during repair of AFW piping.

One or two AFW trains inoperable in MODE 1, 2, or 3 due to inoperable AFW piping affecting the AFW flow path(s) to one steam generator.

G.1 Isolate AFW flow path(s) to affected steam generator.

AND G.2 Restore AFW train(s) to OPERABLE status.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 7 days DIABLO CANYON - UNITS 1 & 2 3.7-11a Unit 1 - Amendment No.

Unit 2 - Amendment No.

236 238

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 236 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 238 TO FACILITY OPERATING LICENSE NO. DPR-82 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2 DOCKET NOS. 50-275 AND 50-323

1.0 INTRODUCTION

By letter dated August 12, 2020 (Reference 1), as supplemented by letters dated August 16, 2020; August 18, 2020; and August 20, 2020 (Reference 2, Reference 3, and Reference 4, respectively), Pacific Gas and Electric Company (PG&E, the licensee) requested changes to the Technical Specifications (TSs) for Diablo Canyon Nuclear Power Plant (Diablo Canyon), Units 1 and 2.

The proposed amendments would avoid an unnecessary plant shutdown during the expected time needed to perform potential repairs to the Unit 1 auxiliary feedwater (AFW) system piping that PG&E conservatively anticipates may be identified during the Diablo Canyon Unit 1, Cycle 22, planned upcoming inspections to the AFW system. Specifically, the proposed amendments would provide a new TS 3.7.5, Auxiliary Feedwater (AFW) System, Condition G, to allow operation of Diablo Canyon Unit 1 for up to 7 days when the AFW system is aligned in a manner for which current TS 3.7.5 would require shutdown. The amendments are only for Cycle 22 during repair of the AFW piping.

As discussed in its license amendment request (LAR) dated August 12, 2020, the licensee requested that the proposed amendments be processed by the U.S. Nuclear Regulatory Commission (NRC, the Commission) on an exigent basis in accordance with the provisions in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.91(a)(6). The NRC staffs evaluation regarding the exigent circumstances is discussed in Section 4.0 of this safety evaluation (SE).

The supplemental letters dated August 16, 2020; August 18, 2020; and August 20, 2020, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the San Luis Obispo News Tribune, located in San Luis Obispo, California on August 16; 2020; August 17, 2020; and August 18, 2020.

2.0 REGULATORY EVALUATION

2.1 AFW System Description Section 6.5, Auxiliary Feedwater System, of the Diablo Canyon Final Safety Analysis Report Update (FSARU) (Reference 5) states:

The AFW system serves as a backup supply of feedwater to the secondary side of the SGs [steam generators] when the main feedwater system is not available, thereby maintaining the heat sink capabilities of the SGs. As an ESFs

[engineered safety features] system, the AFW system is directly relied upon to prevent core damage and RCS [reactor coolant system] overpressurization in the event of transients such as a loss of feedwater or a secondary system pipe rupture, and to provide a means for plant cooldown following any plant transient.

Section 6.5.2, System Description, of the Diablo Canyon FSARU states:

Following a reactor trip, decay heat is dissipated by evaporating water in the SGs and venting the generated steam either to the condensers through the 40 percent steam dump valves or to the atmosphere through the SG safety valves or the 10 percent atmospheric dump valves. SG water inventory must be sufficient to ensure adequate heat transfer and decay heat removal. The AFW system must be capable of functioning for extended periods, allowing time either to restore main feedwater flow or to proceed with an orderly cooldown of the reactor coolant to 350 °F [degrees Fahrenheit] where the RHR [residual heat removal] system can assume the burden of decay heat removal (refer to Section 5.5.6).

AFW system flow and emergency water supply capacity must be sufficient to remove core decay heat, reactor coolant pump heat, and sensible heat during the plant cooldown. The AFW system can also be used to maintain the SG water level above the tubes following a LOCA [loss-of-coolant accident]. The water head in the SGs prevents leakage of fission products from the RCS into the secondary side once the RCS is depressurized.

The AFW system is comprised of three independent pump trains. Two trains consist of motor-driven AFW pumps backed by Class 1E power supplies, and one train consists of a turbine-driven AFW pump with a Class 1E 125-Vdc [volt direct current] steam inlet admission valve.

The motor-driven AFW pumps are each aligned to two SGs. Flow to the SGs is modulated by PG&E Design Class I, electro-hydraulic level control valves (LCVs) powered from Class 1E power supplies. These valves also provide for pump runout protection.

The turbine-driven AFW pump is aligned to all four SGs and contains motor-operated LCVs; however, these valves do not automatically modulate.

AFW is provided to the pumps from the PG&E Design Class I CST [condensate storage tank], which is backed by the PG&E Design Class I fire water storage tank (FWST), and by the PG&E Design Class II raw water storage reservoirs.

The branch connection on two main steam lines for the auxiliary feed pump turbine is provided with isolation valves and CVs [check valves].

In the unlikely event of a complete loss of the preferred power supply and main generator electrical power to the station, decay heat removal would continue to be ensured by the availability of one turbine-driven, and two motor-driven AFW pumps (powered by the standby power source), and steam discharged to atmosphere through the SG power operated relief valves and/or the spring-loaded safety valves. The system is shown in simplified form in Figure 6.5-1.

For the detailed piping schematic, refer to Figure 3.2-3, Sheets 3 and 4.

2.2 Licensee-Proposed Changes Limiting Condition for Operation (LCO) 3.7.5 requires three trains of AFW to be operable in MODE 1 (Power Operation), MODE 2 (Startup), MODE 3 (Hot Standby), and MODE 4 (Hot Shutdown). The LCO in Mode 4 is only applicable when a steam generator (SG) is being relied upon for heat removal. The TS provides Conditions, Required Actions, and Completion Times specifying permissible actions (e.g., continued operation) during the time that LCO 3.7.5 is not met. The allowable time for continued operation varies from 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to 7 days, depending on the applicable condition. The 7-day completion time in Condition A of TS 3.7.5 is limited to inoperability of an AFW turbine train due to one inoperable steam supply or due to the inoperability of the turbine-driven AFW pump during conditions following refueling up to Mode 3.

The licensee requested a new remedial action for the TSs to address a situation where the AFW system was inoperable due only to piping failures. The licensee limited its request to one time for Unit 1 during Cycle 22 during repair of AFW piping. Under the new TS condition, full-power operation could continue for 7 days while repairs are underway. If at the end of 7 days, the AFW system was not returned to service, then the licensee would have to shut down (i.e., exit the modes of applicability) in 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .

Specifically, the proposed amendments would provide new required actions and completion times for a new Condition G to TS 3.7.5 to allow continued operation of Diablo Canyon Unit 1 during Cycle 22 with the AFW system aligned for repairs. The new alignment proposed in Condition G is for one or two AFW trains inoperable in Modes 1, 2, or 3 due to inoperable AFW piping affecting the AFW flow path(s) to one SG. The new Condition G includes required actions to isolate flow to the affected SG within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and to restore the AFW system to OPERABLE status within 7 days. Because the Diablo Canyon TSs are combined for Units 1 and 2, the changes are requested for Diablo Canyon Units 1 and 2. However, the proposed Condition G would include a note specifying that the condition is "only applicable to Unit 1 once during Unit 1 Cycle 22 during repair of AFW piping." TS 3.7.5 Conditions B and D are correspondingly revised to add reference to new Condition G.

2.3 Reason for Proposed Changes Diablo Canyon Unit 2 was manually shutdown and transitioned to Mode 3 on July 17, 2020, due to increasing hydrogen usage in the Unit 2 main generator. While the unit was still in Mode 3, leakage was observed on July 23, 2020, at the elbow downstream of level control valve LCV-111 in the common discharge line from AFW pumps 2-1 and 2-2. Repairs were made at that location and at additional locations where there were no leaks, but pipe wall thickness did not meet minimum American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) requirements. After completion of repairs, Unit 2 was returned to power

operation. The licensee attributed the cause for corrosion of the affected piping to its location in an outside environment and the proximity to the atmospheric dump operation.

The Diablo Canyon Unit 1 piping also contains segments of AFW piping located outdoors. The licensee performed a walkdown of the Unit 1 piping located outdoors on July 25, 2020, and July 26, 2020. The walkdown did not reveal any leaks or areas of immediate concern.

However, the licensee desires to proceed with the additional inspections. A particular area of inspection interest is the AFW piping at the discharge of level control valves LCV-111 (motor-driven AFW pump 1-2) and LCV-107 (turbine-driven AFW pump). The additional inspections could reveal items that need repair. The licensee stated that, based on experience with Unit 2 repairs, potential repairs on Unit 1 systems could require up to 7 days to complete.

The licensee determined that piping repair on the downstream side of valves LCV-111 and LCV-107 could only be performed by isolating SG 1-2, with each of the remaining three SGs (SGs 1-1, 1-3, and 1-4) capable of receiving AFW from two pumps - one turbine-driven and one motor-driven.

The piping section downstream of valves LCV-111 and LCV-107 contains both a check valve and a manual isolation valve that can be used to prevent back flow from SG 1-2. The licensee stated that if repairs are needed, then valves LCV-111 and LCV-107 will be closed to isolate the piping section from the discharge side of the turbine-driven and motor-driven AFW pumps. This configuration will allow the turbine-driven AFW pump 1-1 and motor-driven AFW pump 1-2 to maintain their capability to inject AFW to SG 1-1. The licensee stated that while in this alignment, the turbine-driven AFW pump and both motor-driven AFW pumps will remain available to supply AFW to the remaining three SGS. The licensee provided evaluations to show that with all three AFW pumps and three of the four SGs available, the accident analyses requirements of the AFW system will be maintained during the 7-day completion time The LAR is based on deterministic evaluations supported by risk management actions (RMAs) for defense in depth.

2.4 Regulatory Requirements Under Title 10 of the Code of Federal Regulations (10 CFR 50.90), whenever a holder of a license wishes to amend the license, including TSs in the license, an application for amendment must be filed, fully describing the changes desired. Under 10 CFR 50.92(a), in determining whether an amendment to a license will be issued to the applicant, the Commission will be guided by the considerations which govern the issuance of initial licenses to the extent applicable and appropriate. Under the common standards for licenses and construction permits in 10 CFR 50.40(a), the Commission will be guided by, among other things, whether the operating procedures, the facility and equipment, the use of the facility, and other TSs, collectively, provide reasonable assurance that the applicant will comply with the regulations and that the health and safety of the public will not be endangered. Additionally, the considerations specifically for issuance of operating licenses in 10 CFR 50.57(a)(3) similarly, provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public.

The regulation in 10 CFR 50.36(c)(2)(i) states, in part, that:

Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall

shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

The remedial actions must provide the requisite reasonable assurance of public health and safety.

2.5 Regulatory Guidance For human factors engineering (HFE), the NRC staff considered the following guidance:

NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition. Chapter 18, Revision 3, Human Factors Engineering, dated December 2016 (Reference 6), provides guidance for the review of HFE considerations of plant modifications and important human actions.

NUREG-1764, Guidance for the Review of Changes to Human Actions, Revision 1, dated September 2007 (Reference 7). This document provides guidance for reviewing changes in human actions such as those that are credited in nuclear power plant safety analyses.

NUREG-0711, Human Factors Engineering Program Review Model, Revision 3, dated November 2012 (Reference 8). This document provides guidance for the review of HFE programs to ensure that HFE practices and guidelines are appropriately considered and incorporated in the plant design and modifications.

The NRC staffs guidance for the review of TSs is in Chapter 16.0, Technical Specifications, of NUREG-0800, Revision 3, dated March 2010 (Reference 9). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared standard technical specifications (STS) for each of the LWR nuclear designs. Accordingly, the NRC staffs review includes consideration of whether the proposed changes are consistent with the applicable reference STS (i.e., the current STS), as modified by NRC-approved travelers. The staff also considered the factors in the Commissions Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors (58 FR 39132; July 22, 1993) (describing how the reasons or bases for TSs should address questions like: Why should this remedial action (Action) be taken if the associated LCO cannot be met? How does this Action relate to other Actions associated with the LCO? What justifies continued operation of the system or component at the reduced state from the state specified in the LCO for the allowed time period?). The TSs for Diablo Canyon Units 1 and 2 are based on NUREG-1431, Standard Technical Specifications - Westinghouse Plants, Revision 1.0, Volume 1, Specifications, and Volume 2, Bases, dated April 1995 (Reference 10).

3.0 TECHNICAL EVALUATION

3.1 Background on Piping Corrosion By supplemental letters dated August 16, 2020, and August 18, 2020, the licensee provided a piping corrosion assessment. The NRC staff noted that the material, environment, and aging effect/mechanism identified by the licensee are carbon steel, outdoor air, and general and/or local outside metal loss patterns, respectively. External surfaces of carbon steel piping exposed to outdoor air are susceptible to loss of material due to general, pitting, and crevice corrosion.

Based on information in Corrosion in Marine Atmospheres (Reference 11), the NRC staff noted that atmospheric corrosion rates of carbon steel in a marine environment can be influenced by several factors such as distance from the sea and the length of time exposed to moisture.

The NRC staff noted that following removal of insulation, the licensee will perform visual inspections by trained engineering personnel of pipe surfaces to determine if corrosion product removal is needed. In addition, the NRC staff noted that if corrosion product removal is needed, the licensee will use pit-depth gauges and ultrasonic testing to quantify the extent of corrosion.

3.2 System Safety Analysis Basis Diablo Canyon FSARU Chapter 15, Accident Analysis, broadly divides the analyses into four categories.

Condition I - Normal Operation and Operational Transients (Initial Conditions)

Condition II - Faults of Moderate Frequency Condition III - Infrequent Faults Condition IV - Limiting Faults 3.2.1 Review of Condition 1 Events There were no Condition I analyses that required AFW.

3.2.2 Review of Condition II Events The following Condition II accident analyses were reviewed by the NRC staff.

Loss of Feedwater Transient During a loss of normal feedwater event, the reactor will trip, and two motor-driven AFW pumps and one turbine-driven AFW feedwater pump will start. The motor-driven pumps are connected to Class 1E buses and are supplied by the diesels if a loss-of-offsite power (LOOP) occurs. The turbine-driven pump utilizes steam from the secondary system and exhausts it to the atmosphere. FSARU Section 15.2.8 addresses the loss of feedwater event. The transient is modeled with an assumed single failure of the turbine-driven pump, resulting in the remaining two motor-driven pumps operable and feeding all four SGs with a total of 600 gallons per minute (gpm). Based on FSARU Section 15.2.8.3, the total of 600 gpm is an assumption used in the analysis.

The licensee stated that isolation of SG 1-2 during the system alignment addressed by Condition G means that the two available motor-driven pumps can only pump to three SGs, but the pumping rate will still be over 600 gpm. The staff notes that this rate is consistent with the assumptions in the FSARU.

FSARU Section 15.2.8 analysis shows that the AFW system is capable of removing the stored energy and residual decay heat and reactor coolant pump (RCP) heat, thus preventing either over-pressurization of the RCS or liquid relief through the pressurizer power-operated relief valves or safety valves. FSARU Chapter 6 also contains a loss of normal feedwater transient in Section 6.5.3.7, termed better estimate analysis, which was performed for AFW reliability demonstration. FSAR Section 6.5.3.7 notes that the FSAR Section 15.2.8 analysis has

considerable margin when four SGs are credited and that the better estimate analysis shows successful event mitigation with just two SGs receiving a total of 390 gpm.

The licensee stated that each motor-driven pump provides 100 percent of the feedwater flow required for removal of decay heat from the reactor based on better estimate conditions and that the better estimate evaluation provides a reliability basis for assuming availability of both motor-driven AFW pumps for accident analysis. The licensee stated that the dual approach has been used at other Westinghouse sites such as the Callaway Plant, Unit 1 (Reference 12).

The NRC staff compared the information in the application with the assumptions and analyses in the FSARU. In particular, the staff considered if operating in the system alignment taken under proposed TS 3.7.5 Condition G would be inconsistent with the FSARU. As discussed above, the anticipated AFW flow rate during the loss of AFW transient is above 600 gpm, which is well above the 390 gpm rate that has been shown to be successful. Therefore, the NRC staff finds that the loss of feedwater transient evaluated in the FSARU can be successfully mitigated during proposed Condition G, with SG 1-2 isolated from AFW.

LOOP to the Station Auxiliaries During a complete LOOP and a turbine trip, there will be loss of power to the plant auxiliaries (e.g., RCPs, condensate pumps), and the AFW system is started automatically. Upon the loss of power to the RCPs, coolant flow necessary for core cooling and the removal of residual heat is maintained by natural circulation in the reactor coolant loops.

FSARU Section 15.2.9 assumes AFW flow to all four SGs. For the proposed Condition G, isolation of SG 1-2 will result in AFW flow to only three SGs. Without the RCPs operating, the energy that needs to be removed by the AFW is reduced by 20 megawatts (MW) when compared to the loss of normal feedwater event. The staff determined that the LOOP case is bounded by loss of normal feedwater event discussed in the previous section. Accordingly, for the same reasons discussed in the previous section, the NRC staff finds that the LOOP to the station auxiliaries can be successfully mitigated during proposed Condition G, with SG 1-2 isolated from AFW.

3.2.3 Review of Condition III and IV Events In the event that leaks are discovered in AFW piping downstream of valves LCV-111 and LCV-107, SG 1-2 will be isolated to perform repairs, with all three AFW pumps (one turbine-driven and two motor-driven) available to deliver AFW to the remaining three SGs. This is a less-limiting situation than what was assumed in the FSARU accident analysis for Condition III and IV events, which assume flow from only one motor-driven pump to two SGs. Condition III and IV analyses address minor breaks and major ruptures, respectively, in primary systems (e.g., RCS) and secondary systems such as main steam lines, main feedwater lines, and steam generator tube rupture (SGTR).

The staff notes that in the proposed TS 3.7.5 Condition G action, there would still be three SGs available. As discussed above, only two SGs are assumed for Condition III and IV events.

Therefore, AFW contribution to mitigation of Condition III and IV events is acceptable because of the availability of three SGs, compared to the FSARU analyses requirement of two SGs.

Some of the secondary pipe ruptures, depending on the rupture location, could cause diversion of the AFW flow out of the break. This is a recognized concern, and the operators are trained to

follow procedures and isolate the rupture within 10 minutes of the break, as described in the FSARU, which is a requirement that would continue to apply when the plant is in proposed Condition G. The associated HFE evaluation is in Section 3.3 of this SE.

The SGTR accident analysis is presented in FSARU Section 15.4.3. The analysis included consideration of AFW flow injection to a faulted SG (i.e., with a SGTR). Successful mitigation of the accident requires operator actions (e.g., identification and isolation of the ruptured SG, cooldown and depressurization of the RCS to restore inventory, and termination of safety injection to stop primary to secondary leakage). The analysis states that AFW injection combined with primary-to-secondary leakage in a faulted SG results in a faster rise in that SG level, which helps operators to diagnose the faulted SG. During proposed Condition G, when SG 1-2 is isolated, the capability to inject AFW would be lost. In response to a request from the NRC staff by letter dated August 20, 2020, the licensee provided additional information to address the loss of AFW injection capability with respect to identifying a faulted SG and the impact on dose consequences, as summarized below.

Methods Available for Identification of a Ruptured SG Since AFW is isolated to SG 1-2, any rise in SG 1-2 level is considered unexpected. In order to be in emergency operating procedure (EOP) E-3, Steam Generator Tube Rupture, a safety injection signal would have occurred, and the ruptured SG 1-2 would be filling at approximately twice that of the intact SGs from AFW alone.

Radiation monitors (RMs) 71 through 74 (RM-72 is associated with SG 1-2) are available to detect upward trend or spikes. These monitors are conservatively omitted in FSARU Section 15.4.3, but the capability is not impacted by AFW isolation. Also, per guidance in EOP E-3, the SGs can be sampled by chemistry personnel to assist in the identification of the ruptured SG.

The NRC considered how the existing plant equipment and procedures would be used in conjunction with the permission in the proposed new Condition G for identifying a ruptured SG tube. The NRC concluded that sufficient guidance and procedural control (for example, those described above) would allow the licensee to maintain public health and safety. The information provided by the licensee, as discussed above, provides assurance that isolation of SG 1-2 from AFW will have an insignificant impact on operator detection of the faulted SG (i.e., with SGTR),

isolated or unisolated.

Dose Consequences The SGTR thermal hydraulic dose input analysis (FSARU Section 15.5.20) focuses on maximizing the RCS mass release to the environment for a ruptured SG to conservatively estimate the offsite and control room dose. The contribution of mass released to the atmosphere during the approximate 10-minute assumed run time of the AFW pumps is less than 5 percent of the overall mass release during the event. In addition, Unit 1 is currently operating at less than 0.1 percent of the TS-allowed RCS and secondary activity levels.

In addition, the licensee-provided information assures that dose consequences to offsite and the control room are bounded by the existing analysis for the SGTR event.

The NRC staff evaluated the licensees assertions on dose consequences and found them to be reasonable and consistent with the FSARU.

Consideration of the Plants Design and General Design Criteria Diablo Canyon Units 1 and 2 were designed to comply with the Atomic Energy Commission (now the NRC) General Design Criteria (GDC) for Nuclear Power Plant Construction Permits published in July 1967. The NRC staff reviewed the 1967 GDC that were listed by the licensee in the LAR as being applicable to this change. The proposed change adds a new Condition G to TS 3.7.5. In this condition, a single SG will be isolated from AFW for 7 days, but still available, along with the other three SGs, for power generation. During the 7-day completion time afforded by the new Condition G, two segments of AFW piping on the discharge side of valves LCV-111 and LCV-107 would be closed to perform piping repairs that may be needed.

Normally, these two valves are meant for AFW injection and level control in SG 1-2. Since SG 1-2 will be isolated during Condition G, using these valves for an isolation function only during Condition G is acceptable. There are no physical changes related to the location of the piping, changes related to instrumentation and controls, and systems operation that would impact the GDC. The NRC staff review of the proposed changes indicates that the functional requirements of the AFW system will be maintained, except injection capability to SG 1-2 during proposed TS 3.7.5 Condition G. Therefore, the NRC staff concluded that a significant number of GDC listed in the LAR, while applicable to the design of the AFW system, are not applicable to the current limited scope of the changes to the system configuration during the 7-day operation under proposed Condition G of TS 3.7.5. Therefore, the staff limited the review to determine the impact of the temporary configuration of the AFW system during the 7-day period to the following GDC:

GDC Protection System Redundancy and Independence GDC Separation of Protection and Controls Instrumentation GDC Protection Against Multiple Disability of Protection Systems GDC 20, 22, and 23 are related to redundancy, independence, separation, controls, instrumentation, and protection of the protection systems. During Condition G, SG 1-2 will be isolated, and level control valves LCV-111 and LCV-107 will be shut. Since the LCVs are only designed to serve SG 1-2, this change essentially is limited to the isolation of SG 1-2 from the normal configuration. In the LAR, the licensee stated that the isolated SG TS-related equipment will be operational and listed the equipment in the letter dated August 16, 2020 (Reference 2).

The list contains equipment from various TSs, which the licensee stated will be in effect and operational in accordance with its corresponding TSs. The only function lost is AFW injection capability to SG 1-2 during the 7 days of Condition G, but without impacting the ability to bring the plant to a shutdown if circumstances require.

Therefore, the staff concludes that the plant protection system redundancy, dependence, separation, and multiple disability are still maintained. The staff also considers that the engineering safety features design bases and components capability are still maintained.

3.3 Human Factors Engineering 3.3.1 Description and Assessment of the Credited Operator Action In its letter dated August 12, 2020, as supplemented by letter dated August 20, 2020, the licensee described the proposed changes to Diablo Canyon TS 3.7.5. The licensee stated that the proposed changes address issues associated with a specific AFW system configuration, which is conservatively anticipated to potentially be identified during planned inspections of the Unit 1 AFW system during Cycle 22. In its LAR, the licensee stated that it is conducting these

inspections because the Unit 1 piping between valve LCV-107 and SG 1-2 (Line 570), as well as between valve LCV-111 and SG 1-2 (Line 576), contains piping and elbows that potentially may not meet minimum ASME Code thickness requirements and could require repair.

In its August 20, 2020, letter, the licensee provided a description of how the SGTR accident analysis described in Diablo Canyon FSARU Section 15.4.3 would be affected if an SGTR were to occur in the isolated SG while in proposed Condition G of TS 3.7.5. Specifically, the letter considered potential impacts on the operator actions associated with the detection and isolation of a ruptured SG under the circumstances in question.

In accordance with the guidance provided in Chapter 18 of NUREG-0800, the NRC staff used a graded approach to evaluate the HFE considerations related to the changes described in the LAR. Because the licensee submitted a non-risk-informed LAR, the NRC staff used a qualitative approach in determining the risk significance of the proposed change and the corresponding level of review. In accordance with the generic risk categories established in Appendix A to NUREG-1764, the operator actions associated with SG isolation during both faulted and ruptured SG events are classified as potentially risk-important human actions.

Based upon this screening, the staff determined that the scope of the HFE review should consist of evaluating the potential impact of the proposed TS 3.7.5 change on those operator actions associated with SG isolation during both faulted and ruptured SG events. The NRC staff then performed a qualitative assessment in order to determine whether a Level III HFE review was appropriate. Under the screening and review guidance of NUREG-1764, Level III HFE reviews are limited to verifying that Level III is the appropriate level of review. The staff may, however, choose to increase the review scope.

The licensee stated in its August 12, 2020, letter that a feedline break may create a low resistance path, which could divert AFW flow out of the break and away from intact SGs. The licensee stated that this is a recognized concern for the feedline rupture event response.

Furthermore, the licensee also stated the following:

Operators are currently trained on the time-critical operator action of isolating any ruptured SG within 10 minutes as part of the Diablo Canyon Time Critical Operator Action Program.

Plant Procedure OP1.ID2, Time Critical/Sensitive Operator Action, Action 20, is the operator action to isolate the faulted SG within 10 minutes of the break initiation for the AFW, main steam, and main feedwater systems.

Establishing appropriate AFW flow to the unfaulted SGs occurs as part of isolating the faulted SG.

No AFW flow is credited in the accident analysis until 10 minutes.

The worst identified scenario remains bounded by the accident analysis assumptions.

Diablo Canyon FSARU Section 15.4.2, Major Secondary System Pipe Rupture, states that the analyzed secondary system pipe ruptures include the rupture of a main feedwater pipe. Diablo Canyon FSARU Section 15.4.2.2.3, Analysis of Effects and Consequences, states that for a main feedwater pipe rupture, AFW is assumed to be initiated 10 minutes after the trip. Diablo Canyon FSARU Section 15.4.2.4, Major Rupture of a Main Feedwater Pipe for Pressurizer

Filling, states that the consequences of a feedwater line break for pressurizer filling event can be mitigated, in part, through appropriate operator actions. Diablo Canyon FSARU Section 15.4.2.4.3(4) states that an assumption is made that a time-critical operator action is taken within 10 minutes to isolate the faulted SG and direct AFW flow to an intact SG.

Additionally, FSARU Section 15.4.2.4.4(2) states the following:

Similar to the main feedwater pipe rupture analysis discussed in Section 15.4.2.2, the operators are assumed to isolate the faulted SG within 10 minutes after the low-low SG water level setpoint is reached in accordance with operating procedures. This directs all available AFW flow to the intact SGs.

Diablo Canyon FSARU Section 15.4.3.2, Identification of Causes and Accident Description, states that the operator is expected to identify and isolate a ruptured SG on a restricted time scale in order to minimize contamination of the secondary system and ensure termination of a radioactive release. The operator actions for SGTR recovery are stated to be provided in the EOPs. Specifically, the major operator actions listed include, in part, identifying and isolating feedwater to the ruptured SG. Diablo Canyon FSARU Table 15.4-12, Operator Action Times for Design Basis SGTR Analysis, includes an assumption that a ruptured SG is identified and isolated at either 10 minutes after the initiation of the SGTR or when the narrow range SG level reaches 38 percent, based upon whichever time is longer.

The NRC staff concluded that the licensees proposed change does not constitute more than a minor change to potentially risk-important human actions because the change does not result in any changes to the existing manual operator actions and does not introduce any new manual operator actions associated with faulted or ruptured SG isolation. The NRC staff further concluded that the proposed change does not involve a change in automation for the operator actions associated with the isolation of a faulted or ruptured SG, does not involve a change to the human-system interface, and does not involve a change in operator staffing levels or operator qualifications. The NRC staff considered the above-mentioned factors as part of the qualitative assessment and determined that a Level III HFE review is appropriate. While NUREG-1764 allows Level III HFE reviews to be limited to verifying that Level III is the appropriate level of review, the staff may still choose to increase the review scope. In the present case, the staff elected to further consider available information related to operator training and procedures.

3.3.2 Operator Training and Procedures In its letter dated August 12, 2020, the licensee stated that operators are currently trained on the time-critical operator action of isolating any ruptured SG within 10 minutes as part of the Diablo Canyon Time Critical Operator Action Program. The licensee also stated that the operator action to isolate the faulted SG within 10 minutes of the break initiation for the AFW, main steam, and main feedwater systems is included in Plant Procedure OP1.ID2, Time Critical/Sensitive Operator Action, as Action 20. Furthermore, the licensee stated that the actions associated with the isolation of a faulted SG include establishing appropriate AFW flow to the unfaulted SGs.

In its letter dated August 20, 2020, the licensee stated that the isolation of a ruptured SG would be accomplished using EOP E-3, Steam Generator Tube Rupture. Specifically, the licensee noted that EOP E-3 contains the necessary procedural guidance for identifying and isolating a

ruptured SG if an SGTR were to occur in the isolated SG while in proposed Condition G of TS 3.7.5.

The NRC staff finds that the aforementioned operator training and procedures collectively provide reasonable assurance that the applicant will comply with regulations and that the health and safety of the public will not be endangered.

3.4 Risk Insights While this is not a risk-informed LAR, the licensee provided risk insights related to the proposed change in Section 3 of the enclosure to the LAR. Because this is not a risk-informed LAR, the probabilistic risk assessment (PRA) models used by the licensee to derive risk insights were not reviewed by the NRC staff to determine their technical acceptability to support this SE.

However, the NRC staff considered the licensee-provided qualitative risk insights and associated RMAs as clarified in the supplement dated August 16, 2020, to aid the deterministic review of the proposed change. The NRC staff performed an independent assessment using the NRC Diablo Canyon Unit 1 Standardized Plant Analysis Risk (SPAR) model to evaluate the risk insights and risk contribution from the proposed change. The NRC staffs review confirmed the licensee provided qualitative risk insights and supported the traditional engineering conclusions associated with the proposed change, as well as the licensees proposed RMAs.

The risk insights associated with TS 3.7.5 Condition G are based on considerations of internal events, internal floods, internal fires, and seismic risks. These insights include the risk impact is expected to be low when TS 3.7.5 Condition G is in effect for Unit 1 and does not create a special circumstance described in Appendix D, Use of Risk Information in Review of Non-Risk-Informed License Amendment Requests, of NUREG-0800 (Standard Review Plan),

Chapter 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis General Guidance; and the risk impact is low due to availability of the three AFW pumps to provide flow to the unaffected SGs, availability of main feedwater to SGs, and capability to perform feed and bleed for decay heat removal. The RMAs proposed for TS 3.7.5 Condition G are based on these risk insights and support the availability of AFW, feed and bleed, and main feedwater to maintain defense in depth related to system redundancy, independence, and diversity.

Specifically, regarding fire risk, fire PRAs, similar to deterministic post-fire safe shutdown analyses, do not postulate two fires occurring at the same time. The projected growth and spread of the assumed fire are already accounted for in the fire PRA and deterministic fire safe shutdown analyses. Combined with the criteria (in both analyses) that there will not be a second fire, this indicates that fire prevention actions beyond normal practice are neither risk-significant nor needed to provide reasonable assurance of safe shutdown.

Based on the information provided by the licensee and the evaluation above, the NRC staff concludes that the available risk insights are acceptable for the purposes of supporting the deterministic evaluation.

3.5 TS Changes LCO 3.7.5 requires the AFW system to be operable in MODE 1 (Power Operation), MODE 2 (Startup), MODE 3 (Hot Standby), and MODE 4 (Hot Shutdown). The LCO in Mode 4 is only applicable when an SG is being relied upon for heat removal. The licensee requested a new

remedial action (Condition G) for the TSs to address a situation where the AFW system was inoperable due to piping failures. The proposed TS included a note in Condition G stating:

This Condition is only applicable to Unit 1 once during Unit 1 Cycle 22 during repair of AFW piping. Under the new TSs, full-power operation could continue for 7 days while repairs are underway. If at the end of 7 days, the AFW system was not returned to service, then the licensee would have to shut down (i.e., exit the modes of applicability) in 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months . The licensees proposal would also revise existing Conditions B and D. Condition B would be revised to state (additions in bold) One AFW train inoperable in MODE 1, 2, or 3 for reasons other than Condition A or G. The completion time for Condition B is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Without the proposed change, the Condition G allowed time of 7 days would conflict with the 72-hour time in Condition B for one inoperable AFW train. The proposed change ensures that 7-day completion time for Condition G can be applied for one inoperable AFW train. Similar changes are proposed to Condition D to ensure that 7-day completion in new Condition G takes precedence over existing Condition D for two inoperable trains, which requires commencement of shutdown actions.

The NRC staff reviewed the licensees evaluation as explained below.

The licensee explained that if Condition G is entered, then the new Required Action G.1 provides 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for isolating flow to the affected SG, which also prevents the piping segment(s) where repairs may become necessary from being pressurized with AFW, should the AFW pumps start for any reason during the total completion time under Condition G. The licensee stated that the actions include closing level control valves LCV-107 and LCV-111 on the discharge side of the motor-driven AFW pump 1-2 and turbine-driven AFW pump, respectively.

The staff considers that 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for completion of Required Action G.1 is appropriate, because in addition to closing the valves, additional actions to confirm leaktightness and isolation from plant instrumentation would be necessary.

Required Action G.2 provides 7 days completion time. The licensee stated that the 7 days is based on experience gained on Unit 2, which included time to perform the potential repairs and associated post-maintenance inspections and testing. The licensee explained in the application that continued operation of Unit 1 was justified because the plant remains within the bounds of existing design-basis safety analyses. The staff reviewed this assertion by reviewing the plants design basis and accident analyses in the FSARU. The NRC staffs evaluation of the AFW system capability in terms of mitigating postulated accidents is addressed in Section 3.2 of this SE. Based on this evaluation, the NRC staff determined that the combination of three available AFW pumps and three SGs is equivalent or better than the combination of the AFW equipment assumed in the accident analyses presented in the FSARU. The staff determined that taking action to restore the AFW system to operable status within 7 days was appropriate because it provides sufficient time to repair the Unit 1 AFW piping while still assuring the plant is within the bounds of the existing design-basis safety analyses and there is a low probability of a design-basis accident occurring during this period. Therefore, the available AFW pumps and SGs during Required Action G.2 in the proposed Condition G are capable of mitigating the postulated design-basis accidents at Unit 1.

As part of its assessment, the staff considered how this new required action and completion time relate to the other actions within TS 3.7.5. With regard to how this new required action and completion time relate to the other actions within TS 3.7.5, the staff notes that the licensee revised Conditions B and D to add reference to the new Condition G such that the Condition G

Completion Time of 7 days will apply for the Condition G AFW system alignment, and therefore, take preference over Conditions B and D.

In addition, the staff also reviewed the licensee-presented risk insights associated with TS 3.7.5 Condition G based on considerations of internal events, internal floods, internal fires, and seismic risks (see Section 3.4 of the SE). In the case of fire disabling the control valves to the four SGs from the turbine-driven AFW pump during Condition G, three of the SGs would still continue to receive AFW flow from the two motor-driven pumps, which have their own valves. Based on the NRC staffs accident analyses in Section 3.2 of this SE, the plant can safely shut down with the available SGs. The risk impact is low due to the availability of three AFW pumps and three SGs during proposed Condition G, the availability of main feedwater water to all the SGs, and the capability of feed and bleed for decay heat. This availability of main feedwater to the SGs and feed and bleed maintain defense in depth related to system redundancy, independence, and diversity, and therefore, further reduce risk when TS 3.7.5 Condition G is in effect. Accordingly, for the reasons above, the staff concluded that the new Condition G and the associated actions and completion times provided reasonable assurance of public health and safety.

3.6 Technical Evaluation Conclusion

Based on the technical evaluation described above, the NRC staff finds the proposed changes to provide a new TS 3.7.5 Condition G to address a one-time planned Diablo Canyon Unit 1, Cycle 22, AFW system alignment, will have minimal impact on the continued safe operation and safe shutdown capability of the plant. The alignment of the AFW system under the proposed Condition G will meet the requirements of the accident analyses in Diablo Canyon Units 1 and 2 FSARU Chapter 15. The NRC staff did not perform new accident analyses to arrive at this conclusion; rather, the review is based on comparison of the number of AFW pumps required in the FSARU accident analyses with the number of AFW pumps that would be available with Unit 1 in an action statement and that an equivalent or higher AFW injection capability would exist under the proposed Condition G in TS 3.7.5. In the proposed Condition G action statement, three AFW pumps and three SGs will be available during the 7-day completion time.

The licensee provided an evaluation that concluded the proposed Condition G actions assure that the plant remains within the bounds of existing design-basis safety analyses. The NRC staff reviewed the licensees evaluation and found it acceptable. The new Condition G includes required actions to isolate flow to the affected SG within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and to restore the AFW system to OPERABLE status within 7 days. The proposed TS changes are acceptable and will provide reasonable assurance that the activities at issue will not endanger the health and safety of the public. The NRC staff has evaluated the proposed change considering deterministic information and risk insights to make an integrated decision. With the proposed 2-hour and 7-day CTs, there is reasonable assurance that the activities at issue will not endanger the health and safety of the public.

4.0 EXIGENT CIRCUMSTANCES

4.1 Background

The NRCs regulations contain provisions for issuance of amendments when the usual 30-day public comment period cannot be met. These provisions are applicable under exigent circumstances. Consistent with the requirements in 10 CFR 50.91(a)(6), exigent circumstances exist when: (1) a licensee and the NRC must act quickly, (2) time does not permit the NRC to publish a Federal Register notice allowing 30 days for prior public comment, and (3) the NRC

determines that the amendment involves no significant hazards consideration. As discussed in the licensees application dated August 12, 2020, the licensee requested that the proposed amendments be processed by the NRC on an exigent basis.

Under the provisions in 10 CFR 50.91(a)(6), the NRC notifies the public in one of two ways when exigent circumstances exist: (1) by issuing a Federal Register notice providing an opportunity for hearing and allowing at least 2 weeks from the date of the notice for prior public comments; or (2) by using local media to provide reasonable notice to the public in the area surrounding the licensees facility. In this case, the NRC used local media and published a public notice in the San Luis Obispo News Tribune, located in San Luis Obispo, California (https://www.sanluisobispo.com/), a newspaper local to the licensees facility, on August 16, 2020; August 17, 2020; and August 18, 2020.

4.2 The Licensees Basis for Exigency The licensee provided the following information to explain the exigency of the proposed amendments. Because of localized corrosion identified on Diablo Canyon Unit 2 AFW piping during a recent Diablo Canyon Unit 2 maintenance outage, the licensee intends to perform inspections of Diablo Canyon Unit 1 AFW piping in the near term to ensure that Diablo Canyon Unit 1 is not similarly affected. If similar below-minimum pipe wall thicknesses are found in the Unit 1 AFW system piping and elbows that were found in Unit 2, based on the estimated time-to-repair gained from the Unit 2 repair, it is likely that the current TS 3.7.5 Required Actions B.1 or D.1 would result in the required shutdown of Unit 1. The TS 3.7.5 change would avoid an unnecessary plant shutdown during the expected time needed to perform the potential repairs and associated post-maintenance inspections and testing to the Unit 1 AFW system piping. The licensee stated that it has assessed the potential extent of the Unit 1 AFW system piping repairs based on the required repairs for Unit 2 and is making its best efforts to make a timely application and has not created the exigency. Accordingly, the licensee requested approval of this LAR on an exigent basis.

4.3 NRC Staff Conclusion

Based on the above circumstances, the NRC staff finds that the licensee made a timely application for the proposed amendments following identification of the issue. In addition, the NRC staff finds that exigent circumstances exist in that both the licensee and the NRC must act quickly because if they do not, the AFW inspection results may cause a plant shutdown. The NRC staff finds that time does not permit the NRC staff to publish a Federal Register notice allowing 30 days for prior public comment. Based on these findings, and the determination that the amendments involve no significant hazards consideration as discussed in Section 6.0 below, the NRC staff has determined that a need exists for issuance of the license amendments using the exigent provisions of 10 CFR 50.91(a)(6).

5.0 PUBLIC COMMENTS In accordance with 10 CFR 50.91(a)(6), the NRC used local media and published a public notice in the San Luis Obispo News Tribune, located in San Luis Obispo, California (https://www.sanluisobispo.com/), on August 16, 2020; August 17, 2020; and August 18, 2020.

The notice included the NRC staffs proposed no significant hazards consideration determination and provided an opportunity for public comment regarding the NRC staffs proposed no significant hazards consideration determination. The notice requested that any comments be submitted by August 21, 2020.

The NRC received ten public comments regarding the proposed amendments (References 13 through 22). Some of the issues discussed in the public comments were not submitted according to the procedure described in the public notice or did not pertain to the proposed no significant hazards consideration determination, and some of the comments were received after the August, 21, 2020, deadline. Nevertheless, the NRC staff has made its best attempt to address these comments, in addition to the comments received on the NRC staffs proposed no significant hazards consideration determination. A summary of the comments and the NRC staff responses, grouped by issue, are addressed below.

5.1 AFW System Public Comment The NRC received public comments (References 13, 15, and 18) about the importance of the AFW system. In Reference 13, the public commenter stated that the AFW is the only means to cool the reactor, should it trip. The commenter is concerned that without the AFW system, the licensee would be unable to prevent core damage.

In Reference 14, a commenter stated:

PG&E, in writing, concedes that safety-related piping in more than one AFW flow path may not meet minimum code thickness requirements. Yet, their safety evaluation supporting the amendment implicitly assumes that all piping meets or exceeds the minimum code thickness requirements.

NRC Response Section 2.1 of this SE provides a description of the AFW system. The staff agrees that the AFW system is important to plant safety. The licensees application requests isolating AFW to SG 1-2 and AFW will remain functional for the remaining three SGs. This SE documents the NRC staffs review of the licensees LAR. During the 7-day completion time under proposed TS 3.7.5 Condition G (i.e., one SG would be isolated), all three AFW pumps (one steam-driven and two motor-driven) are available for AFW injection to three SGs. In this configuration, the AFW system is capable of providing decay heat cooling, with margin, under all normal and postulated accidents analyzed in the FSARU.

Each SG has two AFW flow paths - one from the steam-driven AFW pump and another from a motor-driven AFW pump. The two AFW feedwater lines associated with SG 1-2 are the only feedwater lines identified as susceptible to external corrosion. The remaining AFW flow paths would be expected to meet or exceed the minimum code thickness requirements.

5.2 AFW Piping Corrosion Public Comment In Reference 16, the commenter discussed pipe wall thinning and attached multiple generic communications (along with Diablo Canyons responses) pertaining to flow-accelerated corrosion. These comments and attachments suggest that Diablo Canyon did not have an effective NRC-required flow-accelerated corrosion program for monitoring pipe wall thinning.

In Reference 17, after being informed by the NRC that the wall thinning was caused by external corrosion, the commenter acknowledged that flow-accelerated corrosion programs would not be applicable to Diablo Canyons current situation. The commenter then questioned how the NRC could be confident that the replacement of piping would really fix the problem because PG&E had not initially disclosed that the wall thinning was caused by external corrosion. In addition, the commenter observed that PG&E had ample reason to suspect that Unit 1 AFW piping would be similarly degraded as the Unit 2 piping.

NRC Response The scope of the LAR is limited to modifying the TSs to provide requirements on what is permissible upon discovery of suspect piping during operations. The effectiveness of corrosion monitoring programs and the performance of the piping repair are beyond the scope of the requested licensing change.

5.3 Reliability Analysis of the AFW System Public Comment In References 17 and 18, public comments discussed the quality of the licensees PRA analysis, such as common-cause failures and various initiating events. In Reference 18, the public commenter stated:

The only common-cause failure of passive components (e.g., pipes, tanks, heat exchangers, etc.) in PG&Es IPE [Individual Plant Examination] affecting the AFW system was the rupture of a main feedwater or AFW pipe that flooded the AFW pump rooms and disabled both of the motor-driven AFW pumps. This potential flooding scenario was one of the three postulated internal flooding events having risk significance. PG&Es no significant hazards analysis for the exigent LAR stated: The AFW System is not an initiator of any design basis accident or event - a statement apparently contrary to their own IPEs analysis.

NRC Response In Reference 17, a public commenter mentioned a reliability study of the AFW system at Diablo Canyon. However, that study pertains to a September 1980 report on AFW reliability (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17095A390) and is not used in the risk evaluation or LAR. However, the NRC staff addressed why the potential increase in pipe rupture frequency due to corrosion does not impact risk in this section of the SE.

The significant hazards consideration in Section 4.3 of the LAR states, [t]he AFW System is not an initiator of any design basis accident or event. A design-basis accident is a postulated accident that a nuclear facility must be designed and built to withstand without loss to the structures, systems, and components necessary to ensure public health and safety.

Beyond-design-basis accidents is a term used to discuss accident sequences that are possible but were not fully considered in the design process because they were judged to be too unlikely.

Thus, these beyond-design-basis accidents are not usually analyzed in safety analysis reports but are included in PRA studies, such as the IPE, as requested by the NRC in Generic Letter 88-20, Individual Plant Examination for Severe Accident Vulnerabilities -

10 CFR 50.54(f), dated November 23, 1988 (ADAMS Accession No. ML031470299). The

potential AFW flooding scenario discussed in the public comment is a beyond-design-basis accident considered in the licensees IPE. Therefore, the AFW system is not an initiator of any design-basis accident or event.

While this LAR is not a risk-informed application submitted for consideration under the NRC staff guidance in Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decision on Plant-Specific Changes to the Licensing Basis, the NRC staff considered, as discussed in Section 3.4 of this SE, the qualitative risk insights provided by the licensee and performed an independent SPAR evaluation to provide confidence in the staffs deterministic review. The licensees risk insights and RMAs are based on the Diablo Canyon PRAs for internal events, internal floods, internal fires, and seismic that address beyond-design-basis accidents from internal plant causes (such as equipment failures, internal floods, and fires) and external plant causes (such as earthquakes). It should be noted that the Diablo Canyon PRAs are an update to the IPE to reflect the current as-built, as-operated plant and to meet current NRC standards for an acceptable PRA. Considering these hazards (i.e.,

internal events, internal floods, fires, seismic), the expected risk impact is low when TS 3.7.5 Condition G is in effect for Unit 1, due to defense in depth related to system redundancy, independence, and diversity (e.g., availability of AFW to unaffected SGs, main feedwater to the SGs, and feed and bleed). In addition, the affected AFW lines to SG 1-2 are isolated during the 7-day completion time; therefore, any potential increase in pipe rupture frequency due to corrosion/degradation of the affected AFW line should not impact risk.

Finally, the AFW system is not an initiator in the current PRA analysis.

5.4 Postulated Fire Public Comment In Reference 17, the commenter discussed postulated fire risk if valves are closed to allow replacement of externally corroded piping. The public commenter stated:

PG&Es license amendment request was also silent about the potential risk from taking one AFW flow path out of service for up to 7 days whilst Unit 1 continues operating. For example, Appendix 9.5A (ML17157B366) describes how the plant will survive postulated fires in various fire areas. For example, Section 4.2.1 of App. 9-5A reports that a fire in Fire Area 5-A-2 could disable AFW valves LCV-106, LCV-107, LCV-108, and LCV-109.

In Reference 18, the commenter discussed that assessing postulated fire risk regarding a single failure assumption is not required. The public commenter stated:

Unlike the response to other design bases events, the response to a postulated fire need not assume the worst-case single failure. Thus, while the AFW system has redundancy in terms of three pumps and four flow pathways to steam generators for decay heat removal, UFSAR [Updated Final Safety Analysis Report] Appendix 9.5A describes cases where the fire takes away all but a single AFW flow pathway.

NRC Response While this is not a risk-informed LAR submitted for consideration with the NRC staff guidance in Regulatory Guide 1.174, the staff considered, as discussed in Section 3.4 of this SE, the qualitative risk insights provided by the licensee. The licensees risk insights and RMAs are based on the Diablo Canyon PRAs for internal events, internal floods, internal fires, and seismic that address the beyond-design-basis accidents as a result of internal plant causes (such as equipment failures, internal floods, and fires) and external plant causes (such as earthquakes).

The risk insights developed using PRAs consider operation of the plant because PRAs are developed based on the plant operating at full power. Fire PRAs address accident sequences associated with internal plant fires that might occur during power operations. The fire PRAs account for fire-induced initiating events combined with detection and suppression, fire damage to diverse and redundant trains of core cooling equipment, random failures of these equipment, circuits (e.g., spurious actuations), and fire-specific, as well as non-fire-system safety analysis basis related human failures associated with safe shutdown that can lead to undesired consequences (e.g., core damage or large early release). Considering internal fires, including the likelihood of these fires, the expected risk impact is low when TS 3.7.5 Condition G is in effect for Unit 1, due to defense in depth related to system redundancy, independence, and diversity (e.g., availability of AFW to unaffected SGs, main feedwater to the SGs, and feed and bleed).

Regarding AFW flow control valves LCV-106, LCV-107, LCV-108, and LCV-109, these valves control the flow from the turbine-driven AFW pump to one of the four SGs. In the case of a fire disabling these valves, AFW is available to the SGs from the motor-driven AFW pumps, which have their own valves that would still be available. For the proposed fire in Fire Area 5-A-2 occurring during the extended CT, flow would be available to three of the four SGs. There is no fire that results in AFW flow from a single pump to a single SG being required for safe shutdown. In each fire area, flow from at least one AFW pump is available to at least two SGs.

This is accounted for in the licensees risk insights because it is in its fire PRA. In addition, availability of main feedwater to the SGs and feed and bleed maintain defense in depth related to system redundancy, independence, and diversity, therefore, further reducing risk when TS 3.7.5 Condition G is in effect for Unit 1.

5.5 Oversight and Operability Determination Public Comment In Reference 17, the public commenter expressed concern that external corrosion of the AFW piping would be an unreviewed safety question and this degradation represents a new and unanalyzed failure mode for this vital safety system. The commenter also expressed concern that if no new failure mode was introduced, it would involve the increased likelihood that a previously analyzed and reviewed scenario occurs.

In Reference 17, the commenter also asked what might happen if an earthquake occurs while Unit 1 operates with degraded AFW piping.

In Reference 14, a commenter proposed a full-flow evaluation of the degraded AFW piping in Unit 2 in the as-found condition. The commenter stated in Reference 14:

If I was an NRC reviewer, Id feel more comfortable approving this amendment request had PG&E included in its safety analysis an evaluation of the as-found

piping condition on Unit 2 had the AFW system autostarted following a shutdown from full power with one of the three flow paths isolated. If the higher than normal pressure in the three unisolated lines would likely not have failed the as-found thinned piping, theres some reason to believe that Unit 1 would survive such an event as it replaces thinned pipes.

In Reference 18, a commenter questions whether the replaced AFW piping sections on Unit 2 were monitored under PG&Es pipe monitoring program mandated by the NRC, and if so, whether PG&Es failure to adequately implement a pipe monitoring program justifies providing PG&E with a longer time to repair AFW piping on Unit 1.

NRC Response The scope of the LAR is limited to modifying the TSs to provide requirements on what is permissible upon discovery of suspect piping during operations. The effectiveness of a pipe monitoring program is beyond the scope of the requested licensing change. The time for continued operation while a repair is being performed is based on reasonable assurance of public health and safety. Similarly, whether the degradation method is or represents an unresolved safety question is separate and distinct from the request to allow additional time to operate the plant while a repair is being made and the piping is inoperable.

Nuclear power plants in the United States require significant quantities of water for various cooling functions. The water may be salt, brackish, or fresh; and the source of the water may be the ocean, estuary, river, lake, or man-made reservoir. Corrosion in piping is not a new failure mode. For example, some components that support plant operation are located outside and are exposed to known-corrosion-causing effects like snow, ice, rain, fog, and humidity.

Therefore, the licensee has to conduct periodic maintenance to repair and replace piping affected by corrosion.

The AFW system is designed to incorporate defense in depth and redundancy elements to meet the licensing basis of the facility. The feedwater piping associated with SG 1-2 has been identified as susceptible to external corrosion; the licensee plans additional inspection to confirm the AFW piping condition. The AFW piping of concern is located outdoors and is not associated with event initiation. Therefore, no increase in the likelihood of any previously analyzed events would be expected.

A realistic seismic evaluation should be based on actual AFW piping conditions. The licensee has not reported the condition of the AFW piping on Unit 1. Based on the planned additional inspections if a seismic concern is identified by the licensee, a seismic analysis may be performed.

In response to the suggestion that an analysis of the Unit 2 AFW piping as-found configuration should be performed to support Unit 1 operability, this suggestion has been shared with the regional inspection staff.

In Reference 18, the commenter stated that PG&E failed to adequately implement a pipe monitoring program mandated by the NRC, and therefore, should not be given longer time to repair AFW piping on Unit 1. All aspects of the AFW piping integrity issue are subject to NRC oversight, including operability evaluations, corrective actions, extent of condition reviews, and AFW piping monitoring programs. If a performance deficiency is identified, the performance

deficiency will be assessed using the reactor oversight program, and any enforcement action will be consistent with the enforcement policy.

5.6 Davis-Besse Operating Experience Public Comment In Reference 21, a public commenter discussed the significant corrosion experienced in the Davis-Besse Nuclear Power Station (Davis-Besse) reactor vessel head in 2002 and compared that with corrosion of the Diablo Canyon AFW piping. Before identifying the significant corrosion, the Davis-Besse licensee requested, and the NRC approved, a delay in the Davis-Besse vessel head inspection. Similarly, PG&E is delaying its AFW piping inspection.

The commenter stated:

Please ensure the NRC staff does not replicate that mistake by approving the request to operate Diablo Canyon Unit 1 past safety deadlines. The NRC should instead insist the PG&E inspect the Unit 1 AFW piping.

NRC Response In Reference 21, the commenter provided insights comparing the historical sequence of events regarding the corrosion-related degradation of the Davis-Besse reactor head with the corrosion-related degradation affecting the Diablo Canyon AFW piping. The staff notes that the potential impact of the Davis-Besse reactor head degradation could have led directly to a loss of coolant accident initiating event. In contrast, the potential impact of significant degradation of Diablo Canyon AFW piping supplying SG 1-2 would not create an initiating event but could lead to a partial AFW train loss of a mitigating system. Based on its review of the comments, the NRC staff has determined that no changes to the NRC staffs evaluation or conclusions in this SE are needed.

5.7 Exigent Circumstances Public Comment In Reference 22, the commenter provides four examples prior to PG&Es request where the NRC has been asked to extend the 72-hour LCO for one AFW train inoperable and asserts that the NRC has never granted such relief unless an AFW train was declared inoperable. Because PG&E has not declared any AFW train on Diablo Canyon Unit 1 to be inoperable, the commenter states that PG&E should submit a regular LAR with all the attendant public notifications.

NRC Response The NRC staff reviewed the evaluation in Reference 22, including the four examples. The Indian Point example involved an AFW pump with high vibrations but still operable. The licensee requested an exigent amendment to extend the 72-hour LCO to support an evaluation of the AFW pump, repair, and other corrective actions. The regulation in 10 CFR 50.91(a)(6)(vi) requires the licensee to explain the exigency. The NRC staff determined that the Indian Point licensee did not sufficiently explain the exigency associated with high AFW pump vibrations, and therefore, the request to use exigent procedures was denied.

The Catawba example involved a coupling failure of a service water pump that provides cooling for an AFW pump. The plant was operating at the time. The NRC approved a request for an emergency amendment to extend the 72-hour LCO to prevent an unnecessary shutdown. The South Texas Project example involved a loss of AFW instrumentation and controls due to an electrical failure. The plant was operating at the time. The NRC approved a notice of enforcement discretion consistent with its Enforcement Policy for repair of failed AFW control circuitry to prevent an unnecessary shutdown.

The Palisades example involved degraded backup steam supply piping for an AFW turbine.

The plant was shut down at the time. The NRC approved a request for an exigent amendment to remove the backup AFW steam supply as a required element to support operability of the AFW system. The NRC staff determined that the exigency criteria in 10 CFR 50.91 was met because without the amendment, the facility would not have been able to restart.

Unlike the Indian Point example, the staff determined, in Section 4.0 of this SE, that the Diablo Canyon licensee provided sufficient information to justify the use of the exigent provisions in 10 CFR 50.91. Specifically, the licensee requested exigency to support near-term inspections and potential repairs of Unit 1 AFW piping in response to localized corrosion identified on Unit 2 AFW piping. The potential repair activities could cause an unnecessary shutdown. Based on the Catawba, South Texas Project, and Palisades examples, the commenter suggests that inoperability is a necessary factor to use exigent procedures to extend the 72-hour completion time for an AFW train. However the regulation in 10 CFR 50.91(a) contains no such provision.

Accordingly, for the reasons described in Section 4.0 of this SE, the NRC staff has determined that issuance of these license amendments using the exigent provisions of 10 CFR 50.91(a)(6) is appropriate.

6.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

The NRCs regulation in 10 CFR 50.92(c) states that the NRC may make a final determination under the procedures in 10 CFR 50.91 that a license amendment involves no significant hazards consideration if operation of the facility, in accordance with the amendment, would not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

As required by 10 CFR 50.91(a), in its application dated August 12, 2020, the licensee provided its analysis of the issue of no significant hazards consideration, which is presented below:

1.

Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change revises the requirements in Technical Specification (TS) 3.7.5, Auxiliary Feedwater (AFW) System, to provide a new Condition G to address a one-time planned Unit 1 Cycle 22 AFW system alignment for which one or two AFW trains are inoperable in MODE 1, 2, or 3 due to inoperable AFW piping affecting the AFW flow path to one steam generator. The AFW System is not an initiator of any design basis accident or event, and therefore the proposed change does not increase the probability of any accident previously evaluated. The AFW System is

used to respond to accidents previously evaluated. The proposed change affects only the actions taken when portions of the AFW System are inoperable and does not affect the design of the AFW System. With the change to TS 3.7.5, adequate AFW cooling flow continues to be provided for accidents previously evaluated and there is no significant impact on accident consequences. No physical changes are made to the plant. The proposed change does not significantly change how the plant would mitigate an accident previously evaluated.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not result in a change in the manner in which the AFW System provides plant protection. The AFW System will continue to remain available to supply water to three of the four steam generators while in the proposed TS 3.7.5 Condition G to remove decay heat and other residual heat by delivering at least the minimum required flow rate to provide adequate cooling. There are no design changes associated with the proposed changes. The changes to the Conditions and Required Actions do not change any existing accident scenarios, nor create any new or different accident scenarios. The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed). The required manual control of one or two AFW level control valves is not an accident initiator.

Therefore, the proposed change does not create the possibility of a new or different accident from any accident previously evaluated.

3.

Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not impacted by this change. The proposed change will not result in plant operation in a configuration outside the design basis since AFW cooling flow to two steam generators can provide adequate core cooling.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

The NRC staff reviewed the licensees no significant hazards consideration analysis. Based on this review and on the NRC staffs evaluation of the underlying LAR as discussed above, the NRC staff concludes that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the

NRC staff makes a final determination that no significant hazards consideration is involved for the proposed amendments and that the amendments should be issued as allowed by the criteria contained in 10 CFR 50.91.

7.0 STATE CONSULTATION

In accordance with the Commissions regulations, the California State official was notified of the proposed issuance of the amendments on August 12, 2020. The State official provided the following comment (Reference 23):

The California State Liaison Officer David Hochschild does not oppose the Diablo Canyon Power Plant License Amendment Request (LAR) 20-01 Exigent Request for Revision to Technical Specification 3.7.5, Auxiliary Feedwater System. The State Liaison Officer supports activities that enhance and promote safety. The proposed Exigent LAR addresses a potential safety concern but this should not become the normal process. The State Liaison Officer expects both the plant operator and Nuclear Regulatory Commission staff to prioritize and maximize safety.

The NRC staff agrees that the exigent license amendment process is not the normal process.

There are specific requirements for exigent circumstances in 10 CFR 50.91(a)(6).

8.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. 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 made a final determination that no significant hazards consideration is involved for the proposed amendments as discussed above in Section 6.0 of this SE. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth 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.

9.0 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 Commissions 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.

10.0 REFERENCES

1.

Gerfen, P., Pacific Gas & Electric Company, letter to U.S. Nuclear Regulatory Commission, Diablo Canyon Units 1 and 2, Docket No. 50-275, OL-DPR-80, Docket No. 50-323, OL-DPR-82, License Amendment Request 20-01, Exigent Request for Revision to Technical Specification 3.7.5, Auxiliary Feedwater System, dated

August 12, 2020 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML20225A303).

2.

Gerfen, P., Pacific Gas & Electric Company, letter to U.S. Nuclear Regulatory Commission, Docket No. 50-275, OL-DPR-80, Docket No. 50-323, OL-DPR-82, Diablo Canyon Units 1 and 2, Response to NRC Request for Additional Information Regarding License Amendment Request 20-01, Exigent Request for Revision to Technical Specification 3.7.5, Auxiliary Feedwater System, dated August 16, 2020 (ADAMS Accession No. ML20229A016).

3.

Gerfen, P., Pacific Gas & Electric Company, letter to U.S. Nuclear Regulatory Commission, Docket No. 50-275, OL-DPR-80, Docket No. 50-323, OL-DPR-82, Diablo Canyon Units 1 and 2, Response to NRC Request for Additional Information Regarding License Amendment Request 20-01, Exigent Request for Revision to Technical Specification 3.7.5, Auxiliary Feedwater System, dated August 18, 2020 (ADAMS Accession No. ML20231A838).

4.

Gerfen, P., Pacific Gas & Electric Company, letter to U.S. Nuclear Regulatory Commission, Docket No. 50-275, OL-DPR-80, Docket No. 50-323, OL-DPR-82, Diablo Canyon Units 1 and 2, Response to NRC Request for Additional Information Regarding License Amendment Request 20-01, Exigent Request for Revision to Technical Specification 3.7.5, Auxiliary Feedwater System, dated August 20, 2020 (ADAMS Accession No. ML20233B187).

5.

Pacific Gas and Electric Company, Diablo Canyon Power Plant, Units 1 and 2, Re-Submittal of Revision 24 to Updated Final Safety Analysis Report, dated September 2018 (ADAMS Package Accession No. ML19231A071).

6.

U.S. Nuclear Regulatory, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition, NUREG-0800 Chapter 18, Revision 3, Human Factors Engineering, dated December 2016 (ADAMS Accession No. ML16125A114).

7.

U.S. Nuclear Regulatory Commission, Guidance for the Review of Changes to Human Actions, NUREG-1764, Revision 1, dated September 2007 (ADAMS Accession No. ML072640413).

8.

U.S. Nuclear Regulatory Commission, Human Factors Engineering Program Review Model, NUREG-0711, Revision 3, dated November 2012 (ADAMS Accession No. ML12324A013).

9.

U.S. Nuclear Regulatory Commission, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition, NUREG-0800 Chapter 16, Revision 3, Technical Specifications, dated March 2010 (ADAMS Accession No. ML100351425).

10.

U.S. Nuclear Regulatory Commission, Standard Technical Specifications -

Westinghouse Plants, NUREG-1431, Revision 1.0, Volume 1, Specifications, and Volume 2, Bases, dated April 1995 (ADAMS Accession Nos. ML13196A405 and ML13196A409, respectively).

11.

Richard B. Griffin, Corrosion in Marine Atmospheres, Corrosion: Environments and Industries, Volume 13C, ASM Handbook, edited by Stephen D. Cramer, Bernard S.

Covino, Jr., ASM International, p 42-60, 2006.

12.

Donohew, J., U.S. Nuclear Regulatory Commission, letter to Charles D. Naslund, Union Electric Company, Callaway Plant, Unit 1 - Issuance of Amendment Regarding the Steam Generator Replacement Project (TAC No. MC4437), dated September 29, 2005 (ADAMS Accession No. ML052570054).

13.

Public Comment, email from Paul Blanch dated August 17, 2020 (ADAMS Accession No. ML20232D058).

14.

Public Comment, email from David Lochbaum dated August 17, 2020 (ADAMS Accession No. ML20232C648).

15.

Public Comment, e-mail from Paul Blanch dated August 17, 2020 (ADAMS Accession No. ML20232B759).

16.

Public Comments, e-mail from David Lochbaum dated August 18, 2020 (ADAMS Accession No. ML20233A353).

17.

Public Comment, e-mail from David Lochbaum dated August 19, 2020 (ADAMS Accession No. ML20232D223).

18.

Public Comment on the San Luis Obispo News Tribune, e-mail from Diane Curran dated August 21, 2020 (ADAMS Accession No. ML20235P184).

19.

Public Comment on the San Luis Obispo News Tribune, e-mail from Diane Curran dated August 24, 2020 (ADAMS Accession No. ML20239A771).

20.

Public Comment on the San Luis Obispo News Tribune, e-mail from Diane Curran dated August 24, 2020 (ADAMS Accession No. ML20239A729).

21.

Public Comment e-mail from David Lochbaum to the U.S. Nuclear Regulatory Commission, Diablo Canyon safety issue, dated August 23, 2020 (ADAMS Accession No. ML20240A230).

22.

Public Comment e-mail from David Lochbaum to the U.S Nuclear Regulatory Commission, Diablo Canyon Unit 1 AFW system piping, dated August 26, 2020 (ADAMS Accession No. ML20239A951).

23.

Cochran, J., California Energy Commission, e-mail to Samson Lee, U.S. Nuclear Regulatory Commission, Re: State comments on Diablo Canyon exigent license amendment request, dated August 22, 2020 (ADAMS Accession No ML20235Q275).

Principal Contributors: B. Allik N. Chien J. Hickey T. Hilsmeier N. Karipineni C. Moulton C. Newport A. Russell J. Seymour Date: August 31, 2020

ML20235R635

by e-mail OFFICE NRR/DORL/LPL4/PM*

NRR/DORL/LPL4/LA*

NRR/DSS/STSB/BC(A)* NRR/DSS/SCPB/BC*

NAME SLee PBlechman CTilton BWittick DATE 8/24/2020 8/23/2020 8/27/2020 8/27/2020 OFFICE NRR/DRA/APLB/BC(A)* NRR/DNRL/NCSG/BC* NRR/DRO/IOLB/BC*

OGC - NLO*

NAME JBorromeo SBloom CCowdrey AGhoshNaber DATE 8/25/2020 8/26/2020 8/25/2020 8/28/2020 OFFICE NRR/DORL/LPL4/BC*

NRR/DORL/LPL4/PM*

NAME JDixon-Herrity SLee DATE 8/31/2020 8/31/2020

v t e Letter Sequence Approval
EPID:L-2020-LLA-0176, License Amendment Request 20-01 Exigent Request for Revision to Technical Specification 3.7.5, Auxiliary Feedwater System (Approved, Closed)
Initiation Request, Request, Request Acceptance... Administration Meeting Questions 14 August 2020 Answers 16 August 2020 18 August 2020 Results Approval Other: ML20232C648, ML20233A859, ML20239A728, ML20239A775, ML20239A956, ML20245E312, ML20245E395
MONTHYEAR2020-08-19 [Table View]

ML20235R635

Text

SUBJECT:

Dear Mr. Welsch:

Enclosures:

Attachment:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

3.6 Technical Evaluation Conclusion

4.0 EXIGENT CIRCUMSTANCES

4.1 Background

4.3 NRC Staff Conclusion

6.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

7.0 STATE CONSULTATION

8.0 ENVIRONMENTAL CONSIDERATION

9.0 CONCLUSION

10.0 REFERENCES

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