ML21188A345
| ML21188A345 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 07/08/2021 |
| From: | Samson Lee Plant Licensing Branch IV |
| To: | Welsch J Pacific Gas & Electric Co |
| Lee S, 301-415-3158 | |
| References | |
| EPID L- 2021-LLA-0123 | |
| Download: ML21188A345 (19) | |
Text
July 8, 2021 Mr. James M. Welsch Senior Vice President, Generation and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 56, Mail Code 104/6 Avila Beach, CA 93424
SUBJECT:
DIABLO CANYON NUCLEAR POWER PLANT, UNIT 1 - ISSUANCE OF AMENDMENT NO. 238 RE: REVISION TO TECHNICAL SPECIFICATION 3.7.8, AUXILIARY SALTWATER (ASW) SYSTEM (EMERGENCY CIRCUMSTANCES) (EPID L-2021-LLA-0123)
Dear Mr. Welsch:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 238 to Facility Operating License No. DPR-80 for the Diablo Canyon Nuclear Power Plant, Unit 1. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated July 7, 2021, as supplemented by letter dated July 7, 2021.
The amendment provides a new TS 3.7.8 Condition A note to allow a one-time Completion Time (CT) of 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> to replace the Auxiliary Saltwater System Pump 1-1 motor during Cycle 23.
This license amendment is issued under emergency circumstances as provided in the provisions of paragraph 50.91(a)(5) of Title 10 of the Code of Federal Regulations due to the time-critical nature of the amendment. In this instance, an emergency situation exists in that the amendment is needed to allow the licensee to avoid a plant shutdown.
A copy of the related safety evaluation is also enclosed. The safety evaluation describes the emergency circumstances under which the amendment is issued and the final no significant hazards determination. Notice of Issuance addressing the final no significant hazards determination and opportunity for a hearing associated with the emergency circumstances will be included in the Commissions monthly 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 No. 50-275
Enclosures:
- 1. Amendment No. 238 to DPR-80
- 2. 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. 238 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 July 7, 2021, as supplemented by letter dated July 7, 2021, 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. 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 upon issuance and shall be implemented within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 Facility Operating License No. DPR-80 And the Technical Specifications Date of Issuance: July 8, 2021 Thomas J.
Wengert Digitally signed by Thomas J. Wengert Date: 2021.07.08 13:36:39 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 238 TO FACILITY OPERATING LICENSE NO. DPR-80 DIABLO CANYON NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-275 Replace the following pages of the Facility Operating License No. DPR-80 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 Nos. DPR-80 REMOVE INSERT 3
3 Technical Specifications REMOVE INSERT 3.7-16 3.7-16
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 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 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;
ASW 3.7.8 DIABLO CANYON - UNITS 1 & 2 3.7 PLANT SYSTEMS 3.7.8 Auxiliary Saltwater (ASW) System LCO 3.7.8 Two ASW trains shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One ASW train inoperable.
A.1
NOTE--------------
Enter applicable Conditions and Required Actions of LCO 3.4.6, RCS Loops - MODE 4, for residual heat removal loops made inoperable by ASW.
NOTE----------
A Completion Time of 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> is applicable for ASW pump 1-1 on a one-time basis, for Unit 1 cycle 23.
Restore ASW train to OPERABLE status 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B.
Required Action and associated Completion Time of Condition A not met.
B.1 Be in MODE 3.
AND B.2
NOTE-----------
LCO 3.0.4.a is not applicable when entering MODE 4.
Be in MODE 4.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 12 hours 3.7-16 Unit 1 - Amendment No. 135, 200, 219, Unit 2 - Amendment No. 135, 201, 221 238
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 238 TO FACILITY OPERATING LICENSE NO. DPR-80 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-275
1.0 INTRODUCTION
By application dated July 7, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21188A214), as supplemented by letter dated July 7, 2021 (ADAMS Accession No. ML21189A001), Pacific Gas and Electric Company (PG&E or the licensee) requested changes to the Technical Specifications (TS) (Appendix A to Facility Operating License No. DPR-80) for the Diablo Canyon Nuclear Power Plant, Unit 1 (Diablo Canyon or DCPP).
The proposed changes would revise the operating license to provide a new TS 3.7.8 Condition A Note to allow a one-time Completion Time (CT) of 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> to replace the Auxiliary Saltwater (ASW) System Pump 1-1 motor during Cycle 23.
2.0 REGULATORY EVALUATION
2.1
System Description
The licensee stated in Section 2 of the license amendment request dated July 7, 2021 (ADAMS Accession No. ML21188A214).
The ASW system provides a heat sink from the Pacific Ocean for the removal of process and operating heat from the component cooling water (CCW) system.
The CCW system then provides cooling to PG&E Design Class I components during all MODES of operation, including a design basis accident (DBA), and also to various non-PG&E Design Class I components during normal operation and shutdown.
The ASW system consists of two 100 percent capacity, PG&E Design Class I, cooling water trains. Each train consists of one 100 percent capacity pump, one CCW heat exchanger, piping, valving, and instrumentation. The pumps are automatically started upon receipt of a safety injection signal or 4 kilovolt (kV) automatic transfer. The normal configuration is for one train operation with the second pump cross-tied in stand-by and the second heat exchanger valved
out-of-service except when the ultimate heat sink temperature is 64 degrees Fahrenheit (°F) or higher; therefore no valve realignment occurs with a safety injection signal.
Cross-tie capability of the ASW system exists between the DCPP units. Manual and remote manual system realignment provides for utilization of the second CCW heat exchanger, for use of the standby pump on the same unit, for cross-tying the standby ASW pump from the opposite unit, and for train separation for long-term cooling. The ASW unit cross-tie valve allows one ASW pump on one unit to supply the CCW heat exchanger(s) on the other unit. In the event of a total loss of ASW in one unit, the capability to cross-tie units ensures the availability of sufficient redundant cooling capacity for the affected unit. If the unit cross-tie capability were used, the unit with no operable ASW train would enter Limiting Condition for Operation (LCO) 3.0.3, and the unit from which ASW was being provided would be in a TS 3.7.8 Action A 72-hour Required Action with the cross-tie then declared inoperable.
The principal PG&E Design Class I function of the ASW system is the removal of decay heat from the reactor via the PG&E Design Class I CCW System. The ASW system satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). Additional information about the design and operation of the ASW system is presented in the Updated Final Safety Analysis Report (UFSAR), Section 9.2.7.
The design basis of the ASW system is for one ASW train, in conjunction with the CCW System and the containment cooling systems, to remove accident generated and core decay heat following a design basis loss-of-coolant accident (LOCA) as discussed in Updated Final Safety Analysis Report (UFSAR),
Section 6.2. The ASW system can be re-configured to maintain the CCW temperature to within its design bases limits. The ASW system is designed to perform its function with a single failure of any active component with or without the loss of offsite power. This assumes a maximum ASW supply temperature of 64 °F occurring simultaneously with maximum heat loads on the system. The ASW system, in conjunction with the CCW System, also cools the unit from residual heat removal (RHR) entry conditions to MODE 5 during normal and post-accident operations. The time required for this evolution is a function of the number of ASW pumps, CCW heat exchangers, and RHR heat exchangers that are operating. One ASW train is sufficient to remove decay heat during subsequent operations in MODES 5 and 6.
2.2 Licensee-Proposed Changes The licensee stated in Section 2 of the LAR that the TS 3.7.8 Condition A CT is revised to add the following Note to the current 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> CT:
A Completion Time of 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> is [a]pplicable for ASW pump 1-1 on a one-time basis, for Unit 1 cycle 23.
The licensee provided a marked-up TS page in Attachment 1 of the LAR and a retyped TS page in Attachment 2 of the LAR.
The licensee provided the corresponding TS Bases changes in Attachment 3 of the LAR for information only and stated these changes will be implemented in accordance with the TS Bases Control Program. The changes to the Bases state:
The 72-hour Completion Time is modified by a Note that allows a one-time Completion Time of 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> for ASW pump 1-1, for Unit 1 cycle 23 to support emergent replacement of the ASW pump 1-1 motor. The one-time Completion Time of 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> is reasonable considering the capabilities of the other ASW Pump 1-2 to perform the heat removal function, the cross-tie capabilities of ASW from Unit 2, the low probability of a design basis accident occurring during this period, and the one-time use of a 144-hour Completion Time.
2.3 Reason for Proposed Changes The licensee stated, in Section 2 of the LAR:
On July 5, 2021 at 1038 PDT [Pacific Daylight Time], with operation of Unit 1 in MODE 1, Control Room Operators attempted to swap ASW pumps and CCW heat exchangers for routine weekly swaps per operations procedure. When ASW Pump 1-1 was started, multiple simultaneous alarms, including a ground alarm, were received in the control room. Prior to this, there were no indications that there were any issues with the ASW pump. A megger of the ASW Pump 1-1 motor leads and pump motor determined that the identified ground is in the pump electric motor and replacement of the motor is necessary to eliminate the identified ground condition.
Extensive actions were performed to isolate the ground to the ASW Pump 1-1 motor including the following:
Create and hang clearances to support determination of location of ground, Installed ground buggy Meggered ASW Pump 1-1 with cable and motor terminations connected Disconnected terminations from the ASW Pump 1-1 Motor Meggered ASW Pump 1-1 Motor separated from cables Meggered cable from switchgear breaker to the disconnected motor terminations Calibration checks on the Overcurrent Relay The following actions are being performed to support replacement of the ASW Pump 1-1 Motor:
Create work package and hang clearances to decouple and remove the motor Create work package to remove ocean intake hatches for motor Prepare new motor for installation Uncouple existing motor Remove hatches at ocean intake to support ASW Pump 1-1 Motor removal Remove existing motor Install replacement motor Perform uncoupled pump test run and balance shot as necessary Align and couple motor Remove clearances for ASW Pump 1-1
Perform coupled test run Perform post-maintenance operational verification testing required to declare ASW Pump 1-1 OPERABLE Operations review of operational verification testing results for operability As a result of the significant number of actions already taken and those that still need to be performed, it is estimated the full TS 3.7.8 Condition A Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> will be required, without any contingency time, to return the ASW Pump 1-1 to OPERABLE status.
Additional contingency actions may be needed to complete the restoration of ASW Pump 1-1 to operable status.
It is expected an additional 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> beyond the 72-hour completion time could be required to address contingency actions that may occur during the return of ASW Pump 1-1 to OPERABLE status. DCPP Units 1 and 2 are currently operating in MODE 1 and ASW Pump 1-2, ASW Pump 2-1, and ASW Pump 2-2 are OPERABLE. The TS 3.7.9 Ultimate Heat Sink (UHS) requirements are currently being met with the UHS (Pacific Ocean) temperature below 64 °F.
In response to a Request for Additional Information, the licensee stated that any one or combination of the below issues could result in additional time beyond the 72-hour completion time to replace the motor and complete the operability verification testing:
Proper function of the motor space heater, requiring diagnose and repair Problems with removal of the coupling half from the existing motor or reinstallation on the new motor Achieving proper alignment of the pump to motor coupling Achieving proper tolerances on the motor seismic restraint Not achieving acceptance criteria of the specified operability verification testing and needing to perform the contingency operability verification testing Problems racking in the 4 kilovolt (kV) motor breaker Elevated vibration during test run warranting installation of balance weight Problems with reinstallation of the termination box on the new motor and re-termination of the 4 kV leads Potential reverse rotation following re-termination due to the nature of 3-phase motors 2.4 Regulatory Requirements Under Title 10 of the Code of Federal Regulations (10 CFR) Section 50.90, Application for amendment of license, construction permit, or early site permit, 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 processes to be performed, 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 authorized by the operating license will not endanger the health and safety of the public.
General Design Criteria (GDC)
Diablo Canyon, Units 1 and 2 were designed to comply with the Atomic Energy Commission (AEC) GDC for Nuclear Power Plant Construction Permits, published in July 1967. PG&E has made subsequent commitments to GDCs issued later, that are discussed in Section 3.1 of the UFSAR. Since the requested change does not involve any change to the ASW system design function and only allows a separate one-time extension to an existing TS CT, controlled by 10 CFR 50.36, Technical specifications, for one train of the ASW system being inoperable, there is no impact on the compliance of the ASW system design with the GDC applicable to the ASW system.
The regulation in 10 CFR 50.36(c)(2)(i) states in part:
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 TS remedial actions must provide the requisite reasonable assurance of public health and safety. With the proposed change to the TS 3.7.8 Condition A CT, the ASW system will continue to be able to perform the heat removal function for a DBA. The proposed 144-hour Completion Time can only be used once for ASW Pump 1-1 during Unit 1 Cycle 23. Therefore, the licensee stated that the requisite reasonable assurance of public health and safety will continue to be provided.
3.0 TECHNICAL EVALUATION
3.1 Technical Specification Changes TS 3.7.8 Condition A currently allows a CT of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The TS 3.7.8 Condition A CT is revised to allow a one-time CT of 144-hours during Unit 1 Cycle 23. The licensee stated that the one-time use of a 144-hour CT provides sufficient time to perform the emergent replacement of the ASW Pump 1-1 motor and the required post-maintenance testing.
TS 3.7.8 Condition A corresponds to a level of degradation in which one train of the ASW system is inoperable and the system has lost its redundancy to perform the heat removal function. The ASW system is designed to perform its function with a single failure of any active component with or without the loss of offsite power. This assumes a maximum ASW temperature of 64 °F occurring simultaneously with maximum heat loads on the system. In the July 7, 2021, supplement, the licensee stated that the intake and seawater temperature will remain below 64 °F for the duration of the proposed 144-hour Completion Time. Therefore, the NRC staff finds that the safety analysis will remain bounding during the proposed extended Completion Time. In addition, cross tie capability of the ASW system exists between the DCPP units. Manual and remote manual system realignment provides for utilization of the second CCW heat exchanger, for use of the standby pump on the same unit, for cross-tying the standby
ASW pump from the opposite unit, and for train separation for long term cooling. The ASW unit cross-tie valve allows one ASW pump on one unit to supply the CCW heat exchanger(s) on the other unit. In the event of a total loss of ASW in Unit 1, the capability to cross-tie units ensures the availability of sufficient redundant cooling capacity for Unit 1. If the Unit 1 ASW Pump 1-2 were also to become inoperable and unit cross-tie capability used, Unit 1 would have no operable ASW train and would enter LCO 3.0.3, and Unit 2 would be in a TS 3.7.8 Condition A 72-hour action with the cross-tie declared inoperable.
The NRC staff finds the one-time CT of 144-hours for the inoperable ASW Pump 1-1 during Cycle 23 acceptable based on the capabilities of ASW Pump 1-2 to perform the design basis heat removal function, the cross-tie capabilities of ASW from Unit 2, and the low probability of a DBA occurring during this period.
The proposed one-time 144-hour CT reasonably avoids a potential forced Unit 1 shutdown due to the current 72-hour CT of TS 3.7.8 Condition A being exceeded during the emergent replacement of the ASW Pump 1-1 Motor. In the event of a Unit 1 shutdown, the electrical grid would lose DCPP Unit 1 as a reliable source of baseload power.
The TS Bases change reflects the change to TS 3.7.8 for the one-time 144-hour CT for ASW Pump 1-1. Although the TS Bases are not part of the TS and are not subject to NRC approval, the NRC staff notes that the TS Bases change will be implemented in accordance with the TS Bases Control Program as part of amendment implementation.
3.2 Risk Insights The licensee stated that this LAR is not a risk-informed LAR. However, for additional information, the licensee assessed the risk of extending the CT of one train of the ASW system out for maintenance using the DCPP PRA model, which includes internal events, internal flooding, fire, and seismic events.
For the licensees probabilistic risk assessment (PRA), the total additional exposure time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> was used. This is based on the proposed extension of the current CT from 72-hours (3-days) to 144-hours (6-days). The incremental conditional core damage probability (ICCDP) and the incremental conditional large early release probability (ICLERP) using the core damage frequency and large early release frequency increase above the baseline plant risk have been calculated. The licensee also performed a sensitivity case in support of providing risk insights.
The sensitivity case evaluated numerical impacts (additional ICCDP and ICLERP) of performing pre-planned online maintenance of various other structures, systems, and components, outside the list of risk management actions, while the ASW Pump 1-1 is in TS 3.7.8 Condition A for the proposed CT of 144-hours. In estimating the increase in risk, the common cause and independent failure probabilities for the ASW pumps were left unchanged in the LAR PRA model. The licensee stated that this modeling decision was made because there is currently not sufficient information regarding the cause of the ASW Pump 1-1 motor ground or extent of condition for the other ASW motors to make a different decision. The results of the sensitivity case show that a significant margin remains.
The results of the licensees sensitivity case demonstrate that ICCDP and ICLERP are below the risk significance criteria of Regulatory Guide 1.177 of 1.0E-06 and 1.0E-07, respectively.
Therefore, the proposed separate one-time extension of TS 3.7.8 Condition A Required Action to 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> (6-days) for emergent maintenance for ASW Pump 1-1 Motor is considered to not be risk significant.
The loss of an ASW system could result in a potential reactor coolant pump (RCP) seal LOCA on loss of RCP seal cooling. However, all reactor coolant pumps in both units have installed the Westinghouse Generation III SHIELD shutdown seal, which will allow additional time to establish cooling to charging pumps and provide seal injection to RCPs. The licensee stated that the following risk management actions have been determined to be appropriate based on evaluation of the insights from the PRA assessment and have been accepted by DCPP Operations to be implemented during the proposed TS 3.7.8 Condition A Required Action 144-hour (6-day) CT:
Protect ASW Pumps 1-2, 2-1, 2-2 and the ASW unit crosstie, Protect both Unit 1 CCW heat exchangers, Protect Charging Pump 1-3, Protect AFW pump trains 1-1, 1-2 and 1-3, Tailboard (pre-job brief) operators on the action to trip the RCPs on a loss of CCW, and Tailboard operators on the action to provide backup firewater cooling to the charging pumps.
The licensee stated that the proposed change has been determined to not be risk significant and it does not:
Substantially increase the likelihood or consequences of accidents that are risk significant but are beyond the design and licensing basis of the plant, Degrade multiple levels of defense, or cornerstones in the reactor oversight process, through plant operations or situations not explicitly considered in the development of the regulations, Significantly reduce the availability or reliability of structures, systems, or components that are risk significant but are not required by regulations, or Involve changes for which the synergistic or cumulative effects could significantly impact risk.
Therefore, the licensee states the proposed change 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 (ADAMS Accession No. ML071700658).
3.3 Electrical Evaluation In the LAR, the licensee stated that a megger of the ASW Pump 1-1 motor leads and pump motor identified that there is a ground in the pump electric motor. Extensive actions were performed to isolate the ground to the ASW Pump 1-1 motor, but a replacement of the motor is necessary to eliminate the identified ground condition. The NRC staff requested the licensee to clarify the type of the replacement motor and provide specifications of both the existing and replacement motor. The NRC staff also requested the licensee to confirm that there is no load change as the result of the motor replacement. In its supplemental letter dated July 7, 2021, the licensee stated that the data for the replacement and existing pump motors are the same. The licensee further stated that there are no load changes that result from the motor replacement.
Based on the information the licensee provided, the NRC staff notes that the licensee would replace the existing motor with the same motor, and there are no changes to the electrical load.
The NRC staff determined that there is no impact to the electrical distribution systems.
Therefore, the NRC staff finds the replacement of the motor acceptable.
3.4 Technical Evaluation Conclusion
The NRC staff reviewed the impact the proposed one-time TS 3.7.8 Condition A extended CT would have on the ASW system. The NRC staff therefore finds the loss of one ASW pump, while in the proposed TS 3.7.8 Condition A Completion Time of 144-hours, remains bounded by the UFSAR accident analyses in that the remaining Operable ASW pump can provide the necessary heat removal function. The NRC staff therefore finds the one-time CT extension from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 144-hours acceptable.
Further, the NRC staff finds the proposed change is not risk significant based on risk insights.
Additionally, the NRC staff finds the risk management actions, to be implemented during the proposed TS 3.7.8 Condition A 144-hour CT, as determined by the licensee, are acceptable.
4.0 EMERGENCY CIRCUMSTANCES The NRCs regulations in 10 CFR 50.91(a)(5) state that where the NRC finds that an emergency situation exists, in that failure to act in a timely way would result in derating or shutdown of a nuclear power plant, or in prevention of either resumption of operation or of increase in power output up to the plants licensed power level, it may issue a license amendment involving no significant hazards consideration without prior notice and opportunity for a hearing or for public comment. In such a situation, the NRC will publish a notice of issuance under 10 CFR 2.106, providing for opportunity for a hearing and for public comment after issuance.
As discussed in the licensees application dated July 7, 2021, the licensee requested that the proposed amendment be processed by the NRC on an emergency basis. Section 2 of the LAR states:
The circumstances requiring this emergency amendment request were not foreseeable. A similar motor identified ground event has not occurred in the past 10 years on any four of the DCPP TS 3.7.8 required ASW pumps. At 1500 PDT on July 5, approximately 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the emergent failure of ASW Pump 1-1, PG&E activated the DCPP outage control center to support repair of the ASW Pump 1-1 on a 24-hour per day basis until the pump is returned to OPERABLE status.
The licensee also states:
The TS 3.7.8 change will avoid an unnecessary plant shutdown during the expected time needed to perform the replacement of the ASW Pump 1-1 motor, and associated post-maintenance inspections and testing.
The NRC staff, including the resident inspector, reviewed the licensees basis for processing the proposed amendment as an emergency amendment (as discussed above) and determines that an emergency situation exists consistent with the provisions in 10 CFR 50.91(a)(5).
Furthermore, the NRC staff determined that: (1) the licensee used its best efforts to make a timely application; (2) the licensee could not reasonably have avoided the situation; and (3) the licensee has not abused the provisions of 10 CFR 50.91(a)(5).
Based on these findings, and the determination that the amendment involves no significant hazards consideration as discussed below, the NRC staff has determined that a valid need exists for issuance of the license amendment using the emergency provisions of 10 CFR 50.91(a)(5).
5.0 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 letter dated July 7, 2021, 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 to the requirements in TS 3.7.8, Auxiliary Saltwater (ASW) System Condition A [is] to allow a one-time Completion Time of 144-hours during Unit 1 Cycle 23 for Auxiliary Saltwater System (ASW) Pump 1-1 for the emergent replacement of the pump motor. The ASW system is not an initiator of any UFSAR Chapter 6 or 15 design basis accident or event, and therefore, the proposed change does not increase the probability of any accident previously evaluated. The ASW system is used to supply cooling water to respond to accidents that have been previously evaluated. The proposed change affects only the time allowed for an ASW system train to be inoperable and does not affect the design of the ASW system. With the change to TS 3.7.8, adequate ASW continues to be provided to perform the heat removal function for accidents previously evaluated and there is no significant impact on accident consequences. 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 accident from any accident previously evaluated?
Response: No.
The proposed change does not result in a change in the manner in which the ASW system provides plant protection. The ASW system will continue to perform the function of heat removal while in the proposed revised TS 3.7.8 Condition A. The change does not involve a physical alteration of the plant that impacts the capability of the ASW system to perform its design function.
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 existing design basis since TS 3.7.8 Condition A already allows one train of the ASW system to be inoperable.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on its review of the licensees evaluation set forth above, the NRC staff concludes that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff has made a final determination that no significant hazards consideration is involved for the proposed amendment and that the amendment should be issued as allowed by the criteria contained in 10 CFR 50.92.
6.0 STATE CONSULTATION
In accordance with the Commissions regulations, the California State official was notified of the proposed issuance of the amendments on July 7, 2021. The State official had no comments.
7.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves 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. Accordingly, the amendment meets 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 amendment.
8.0 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.
Principal Contributors: D. Scully R. Beaton R. Wolfgang J. Circle J. Wilson A. Russell J. Seymour A. Athar N. Khan T. Reed Date: July 8, 2021
ML21188A345 OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DSS/STSB/BC(A) NRR/DSS/SCPB/BC NAME SLee PBlechman NJordan BWittick DATE 07/08/2021 07/08/2021 07/08/2021 07/08/2021 OFFICE NRR/DSS/SNSB/BC NRR/DRA/APLA/BC NRR/DRO/IOLB/TL NRR/DEX/EMIB/BC(A)
NAME SKrepel RPascarelli BGreen ITseng DATE 07/08/2021 07/08/2021 07/07/2021 07/08/2021 OFFICE NRR/DEX/EEEB/BC(A) OGC NRR/DORL/LPL4/BC NRR/DORL/LPL4/PM NAME KNguyen STurk JDixon-Herrity (TWengert for)
SLee DATE 07/08/2021 07/08/2021 07/08/2021 07/08/2021