ML25230A133
| ML25230A133 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  (DPR-080, DPR-082) |
| Issue date: | 08/21/2025 |
| From: | Samson Lee Office of Nuclear Reactor Regulation |
| To: | Gerfen P Pacific Gas & Electric Co |
| Lee S, 301-415-3158 | |
| References | |
| EPID L-2024-LLA-0179 | |
| Download: ML25230A133 (1) | |
Text
August 21, 2025 Ms. Paula Gerfen Senior Vice President, Generation and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 56 Avila Beach, CA 93424
SUBJECT:
DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 253 AND 255 RE: REVISION TO TECHNICAL SPECIFICATION 1.1 AND ADDITION OF TECHNICAL SPECIFICATION 5.5.21 TO USE ONLINE MONITORING METHODOLOGY (EPID L-2024-LLA-0179)
Dear Ms. Gerfen:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 253 to Facility Operating License No. DPR-80 and Amendment No. 255 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 December 31, 2024.
The amendments revised TS 1.1, Definitions, and added TS 5.5.21, Online Monitoring Program, to use online monitoring methodology, which provides controls to determine the need for calibration of transmitters using condition monitoring.
A copy of the related Safety Evaluation is enclosed. Notice of Issuance 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 Nos. 50-275 and 50-323
Enclosures:
- 1. Amendment No. 253 to DPR-80
- 2. Amendment No. 255 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. 253 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 December 31, 2024, 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. 253 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 90 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Tony Nakanishi, 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: August 21, 2025 TONY NAKANISHI Digitally signed by TONY NAKANISHI Date: 2025.08.21 15:16:47 -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. 255 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 December 31, 2024, 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. 255 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 90 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Tony Nakanishi, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Facility Operating License No. DPR-82 and the Technical Specifications Date of Issuance: August 21, 2025 TONY NAKANISHI Digitally signed by TONY NAKANISHI Date: 2025.08.21 15:17:10 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 253 TO FACILITY OPERATING LICENSE NO. DPR-80 AND LICENSE AMENDMENT NO. 255 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 the Facility Operating License Nos. DPR-80 and DPR-82, and Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Facility Operating License No. DPR-80 REMOVE INSERT Facility Operating License No. DPR-82 REMOVE INSERT Technical Specifications REMOVE INSERT 1.1-1 1.1-1 1.1-3a 1.1-3a 1.1-5 1.1-5 5.0-17b 5.0-17b 5.0-17c
Amendment No. 253 (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. 253 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 Companys 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. 255 (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. 255, 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.
Definitions 1.1 DIABLO CANYON - UNITS 1 & 2 Rev 13 Page 1 of 27 Tab_1!0u3r13.doc 1113.1118 1.0 USE AND APPLICATION 1.1 Definitions
NOTE------------------------------------------------------------
The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.
Term Definition ACTIONS ACTUATION LOGIC TEST AXIAL FLUX DIFFERENCE (AFD)
CHANNEL CALIBRATION CHANNEL CHECK ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.
An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state and the verification of the required logic output. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices.
AFD shall be the difference in normalized flux signals between the top and bottom halves of an excore neutron detector.
A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY (excluding transmitters in the TS 5.5.21 Online Monitoring Program). Calibration of instrument channels with resistance temperature detectors (RTD) or thermocouple sensors may consist of an in-place qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping or total channel steps.
A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.
(continued) 1.1-1 Unit 1 - Amendment No. 135, Unit 2 - Amendment No. 135, 253 255
Definitions 1.1 DIABLO CANYON - UNITS 1 & 2 Rev 13 Page 5 of 27 Tab_1!0u3r13.doc 1113.1118 1.1 Definitions (continued)
ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME LEAKAGE The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC (including transmitters in the TS 5.5.21 Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.
LEAKAGE shall be:
a.
Identified LEAKAGE 1.
LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; 2.
LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or (continued) 1.1-3a Unit 1 - Amendment No. 135,155,156,192, 244, Unit 2 - Amendment No. 135,155,156,193, 245, 253 255
Definitions 1.1 DIABLO CANYON - UNITS 1 & 2 Rev 13 Page 7 of 27 Tab_1!0u3r13.doc 1113.1118 1.1 Definitions (continued)
PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
QUADRANT POWER TILT RATIO (QPTR)
RATED THERMAL POWER (RTP)
REACTOR TRIP SYSTEM (RTS) RESPONSE TIME The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, and the power operated relief valve (PORV) lift settings and arming temperature associated with the Low Temperature Overpressurization Protection (LTOP) System, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.6.
QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.
RTP shall be a total reactor core heat transfer rate to the reactor coolant of 3411 MWt for each unit.
The RTS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC (including transmitters in the TS 5.5.21 Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.
SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
a.
All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and b.
In MODES 1 and 2, the fuel and moderator temperatures are changed to the hot zero power temperatures.
(continued) 1.1-5 Unit 1 - Amendment No. 135,143, 170, 244, Unit 2 - Amendment No. 135, 171, 245, 253 255
Programs and Manuals 5.5 DIABLO CANYON - UNITS 1 & 2 Rev 38 Page 19 of 29 Tab_5!0u3r38.doc 1113.1604 5.5 Programs and Manuals (continued) 5.5.20 Risk Informed Completion Time (RICT) Program This program provides controls to calculate a RICT and must be implemented in accordance with NEI 06-09-A, Revision 0, "Risk-Managed Technical Specifications (RMTS) Guidelines."
The program shall include the following:
a.
The RICT may not exceed 30 days; b.
A RICT may only be utilized in MODE 1 and 2; c.
When a RICT is being used, any change to the plant configuration, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.
1.
For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.
2.
For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.
3.
Revising the RICT is not required If the plant configuration change would lower plant risk and would result in a longer RICT.
d.
For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:
1.
Numerically accounting for the increased possibility of CCF in the RICT calculation; or 2.
Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.
e.
The risk assessment approaches and methods shall be acceptable to the NRC.
The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods approved for use with this program, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.
(continued) 5.0-17b Unit 1 - Amendment No. 245, Unit 2 - Amendment No. 247, 253 255
Programs and Manuals 5.5 DIABLO CANYON - UNITS 1 & 2 Rev 38 Page 20 of 29 Tab_5!0u3r38.doc 1113.1604 5.5 Programs and Manuals (continued) 5.5.21 Online Monitoring Program This program provides controls to determine the need for calibration for pressure, level, and flow transmitters using condition monitoring based on drift analysis. It also provides a means for in-situ dynamic response assessment using the noise analysis technique to detect failure modes that are not detectable by drift monitoring.
The Online Monitoring Program must be implemented in accordance with AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters (proprietary version). The program shall include the following elements:
a.
Implementation of online monitoring for transmitters that have been evaluated in accordance with the NRC approved methodology during the plant operating cycle.
1.
Analysis of online monitoring data to identify those transmitters that require a calibration check and those that do not require a calibration
- check, 2.
Performance of online monitoring using noise analysis to assess in-situ dynamic response of transmitters that can affect response time performance, 3.
Calibration checks of identified transmitters no later than during the next scheduled refueling outage, and 4.
Documentation of the results of the online monitoring data analysis.
b.
Performance of a calibration checks of any transmitter where the online monitoring was not implemented during the plant operating cycle no later than during the next scheduled refueling outage.
c.
Performance of calibration checks for transmitter at the specified backstop frequencies.
d.
The provisions of Surveillance Requirement 3.0.3 are applicable to the required calibration checks specified in items a.3, b, and c above.
5.0-17c Unit 1 - Amendment No.
Unit 2 - Amendment No.
253 255
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 253 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 255 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 application dated December 31, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24366A169), Pacific Gas and Electric Company (the licensee) requested changes to the technical specifications (TSs) for the Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Diablo Canyon).
The proposed amendments would allow implementation of an online instrument channel monitoring program by revising TS 1.1, Definitions, in TS section 1.0, Use and Application, and adding a new TS 5.5.21, Online Monitoring Program. The licensee proposes to use the online monitoring (OLM) methodology as the technical basis to change from a time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on OLM results. The proposed changes implement the analytical methodologies described within the topical report (TR) by Analysis and Measurement Services (AMS),
AMS-TR-0720R2-A, Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters (ML21235A493), which was approved by the U.S. Nuclear Regulatory Commission (NRC, the Commission) staff in its email dated August 11, 2021 (enclosed within ML21235A493).
The NRC staffs safety evaluation (SE) approving the AMS OLM TR (Package ML21179A060) states, in part, that the NRC staff finds that implementation of an OLM program in accordance with the approved AMS OLM TR provides an acceptable alternative to periodic manual calibration surveillance requirements upon implementation of the application-specific action items. In its implementation of this OLM program, Diablo Canyon has not proposed any deviations from the methodologies or analyses approved in AMS-TR-0720R2-A.
The NRC staff conducted a virtual regulatory audit between March 10 and August 13, 2025, to examine the licensees non-docketed information on the proposed Diablo Canyon OLM methodology. The NRC staff did not identify any need for additional information and issued an audit summary dated August 19, 2025 (ML25230A241).
2.0 REGULATORY EVALUATION
2.1 Regulations and Guidance The NRC staff considered the following regulatory requirements in reviewing the methodology being implemented in the Diablo Canyon OLM program:
Title 10 of the Code of Federal Regulations (10 CFR) 50.36(c)(1)(ii)(A) states that limiting safety system settings are settings for automatic protective devices related to those variables having significant safety functions. This section requires that where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be chosen so that automatic protective action will correct the abnormal situation before a safety limit is exceeded. It also requires that the licensee take appropriate action and notify the NRC if the licensee determines that an automatic safety system does not function as required. The licensee is then required to review the matter and record the results of the review.
The regulation at 10 CFR 50.36(c)(3) states, Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.
The regulation at 10 CFR 50.36(c)(5) states, in part, that [a]dministrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.
The Diablo Canyon construction permits were issued prior to 1971; therefore, the criteria of the 1967 General Design Criteria (GDC), as documented in the Diablo Canyon Updated Final Safety Analysis Report (UFSAR) (as amended) (Package ML24323A239),
are used as a basis for this SE. The design criteria captured in GDC 13 and 20 are mapped to the following pre-GDCs, 12, 14 and 15.
Criterion 12, 1967 - Instrumentation and Control Systems (Category B), states:
Instrumentation and controls shall be provided as required to monitor and maintain variables within prescribed operating ranges.
Criterion 14, 1967 - Core Protection Systems (Category B), states:
Core protection systems, together with associated equipment, shall be designed to act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits.
Criterion 15, 1967 - Engineered Safety Features Protection Systems (Category B),
states:
Protection systems shall be provided for sensing accident situations and initiating the operation of necessary ESFs [Engineered Safety Features].
The following are the specific NRC guidance documents applicable to the NRC staffs evaluation of the Diablo Canyon OLM program:
NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition, Branch Technical Position (BTP) 7-12, Guidance on Establishing and Maintaining Instrument Setpoints, Revision 6 (ML16019A200).
Regulatory Guide (RG) 1.105, Revision 4, Setpoints for Safety-Related Instrumentation, February 2021 (ML20330A329). This RG describes an approach that is acceptable to the NRC staff to meet regulatory requirements to ensure that:
(a) setpoints for safety-related instrumentation are established to protect nuclear power plant safety and analytical limits, and (b) the maintenance of instrument channels implementing these setpoints ensures they are functioning as required, consistent with the plant TSs. This RG endorses American National Standards Institute (ANSI)/International Society of Automation (ISA) Standard 67.04.01-2018, Setpoints for Nuclear Safety-Related Instrumentation. Among other things, the ANSI/ISA 67.04.01 standard provides criteria for assessing the performance of safety-related instrument channels to ensure they remain capable of achieving their required safety functions in a reliable manner. This performance monitoring process requires the establishment of acceptable As-Found tolerance limits used to check whether an instrument channel is functioning as required, and the establishment of acceptable As-Left tolerance limits used to establish the maximum allowed deviation from the desired setpoint of the instrument channel and still be considered as within calibration.
The following guidance documents provide information associated with the periodic calibration of safety-related instrument channels that was considered by the NRC staff during its evaluation of the Diablo Canyon OLM program:
Generic Letter 91-04, Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle, dated April 2, 1991 (ML031140501), provides guidance on acceptable methods for licensees to justify an increase in calibration surveillance intervals using as-found and as-left calibration data from past calibration surveillances.
Regulatory Issue Summary (RIS) 2006-017, NRC Staff Position on the Requirements of 10 CFR 50.36, Technical Specifications, regarding Limiting Safety System Settings during Periodic Testing and Calibration of Instrument Channels, dated August 24, 2006 (ML051810077), provides regulatory clarification on NRC staff positions in terms of the appropriate determination of TS-related instrument channel operability. The RIS clarifies NRC staff positions about the appropriate establishment of as-found and as-left acceptance tolerances.
2.2 Description of Proposed Changes In its license amendment request (LAR) dated December 31, 2024, the licensee proposed the following specific changes to the TSs for Diablo Canyon.
TS 1.1 CHANNEL CALIBRATION The current CHANNEL CALIBRATION definition in TS 1.1 states:
A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY. Calibration of instrument channels with resistance temperature detectors (RTD) or thermocouple sensors may consist of an in-place qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping or total channel steps.
The revised CHANNEL CALIBRATION definition would state (changes indicated in bold):
A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY (excluding transmitters in the TS 5.5.21 Online Monitoring Program). Calibration of instrument channels with resistance temperature detectors (RTD) or thermocouple sensors may consist of an in-place qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping or total channel steps.
TS 1.1 ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME The current ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME definition in TS 1.1 states:
The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.
The revised ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME definition would state (changes indicated in bold):
The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the
ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC (including transmitters in the TS 5.5.21 Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.
TS 1.1 REACTOR TRIP SYSTEM (RTS) RESPONSE TIME The current REACTOR TRIP SYSTEM (RTS) RESPONSE TIME definition in TS 1.1 states:
The RTS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.
The revised REACTOR TRIP SYSTEM (RTS) RESPONSE TIME definition would state (changes indicated in bold):
The RTS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC (including transmitters in the TS 5.5.21 Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.
TS 5.5.21 Online Monitoring Program There is currently no TS 5.5.21.
The new TS 5.5.21, Online Monitoring Program, would state This program provides controls to determine the need for calibration for pressure, level, and flow transmitters using condition monitoring based on drift analysis. It also provides a means for in-situ dynamic response assessment using the noise analysis technique to detect failure modes that are not detectable by drift monitoring.
The Online Monitoring Program must be implemented in accordance with AMS-TR-0720R2-A, Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters (proprietary version). The program shall include the following elements:
- a.
Implementation of online monitoring for transmitters that have been evaluated in accordance with an NRC approved methodology during the plant operating cycle.
- 1.
Analysis of online monitoring data to identify those transmitters that require a calibration check and those that do not require a calibration check.
- 2.
Performance of online monitoring using noise analysis to assess in-situ dynamic response of transmitters that can affect response time performance.
- 3.
Calibration checks of identified transmitters no later than during the next scheduled refueling outage, and
- 4.
Documentation of the results of the online monitoring data analysis.
- b.
Performance of a calibration check for any transmitter where the online monitoring was not implemented during the plant operating cycle no later than during the next scheduled refueling outage.
- c.
Performance of calibration checks for transmitter at the specified backstop frequencies.
- d.
The provisions of Surveillance Requirement 3.0.3 are applicable to the required calibration checks specified in items a.3, b, and c above
3.0 TECHNICAL EVALUATION
3.1 Description of the OLM Program The proposed Diablo Canyon OLM program is based on the AMS OLM TR, AMS-TR-0720R2-A, which provides a methodology for performing OLM of the output signals of pressure and differential pressure transmitters. This methodology was developed by AMS to be used in nuclear power plants as an analytical tool to measure sensor calibration performance during plant operation between scheduled refueling outages.
3.1.1 OLM Program Implementation The licensee stated in section 3.2 of its LAR, that the AMS Bridge and the AMS Calibration Reduction System software programs were developed under AMSs 10 CFR Part 50, Appendix B, compliant Quality Assurance (QA) program. The NRC staff conducted an inspection of AMS to review AMSs implementation of its QA program with respect to the design, testing, and error controls for the AMS Bridge and the AMS Calibration Reduction System software programs. The NRC staff documented its inspection findings in an inspection report
dated March 14, 2025, Nuclear Regulatory Commission Inspection Report of Analysis and Measurement Services No. 99902075/2025-201 (ML25071A181). As stated in this inspection report, the NRC staff determined that AMS is implementing its design control and test control program in accordance with the regulatory requirements of Criterion III, Design Control, and Criterion XI, Test Control, of Appendix B to 10 CFR Part 50 for the AMS Bridge and the AMS Calibration Reduction System software programs. In addition, the NRC staff reviewed AMSs processes for controlling software errors and with exception of one minor procedural issue, the NRC staff determined that AMS is implementing its non-conforming materials, parts, or components program in accordance with the regulatory requirements of Criterion XV, Nonconforming Materials, Parts, or Components, of Appendix B to 10 CFR Part 50 for the AMS Bridge and the AMS Calibration Reduction System software programs.
3.2 Description and Evaluation of TS Changes The licensees submittal requested approval to implement its OLM program by revising appropriate TS 1.1, Definitions, and adding a new section TS 5.5.21 Online Monitoring Program. The licensee proposes to use the OLM methodology presented in AMS-TR-0720R2-A as the technical basis to change from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on OLM analysis results. Markup of the TS pages were provided in attachment 1 of the enclosure to the LAR dated December 31, 2024.
The regulation at 10 CFR 50.36(a)(1) states, in part, that: [a] summary statement of the bases or reasons for such specifications other than those covering administrative controls shall also be included in the application but shall not become part of the technical specifications. Accordingly, the licensee also submitted for NRC staff information the proposed TS Bases changes that correspond to the proposed TS changes. The NRC staff did not review the TS Bases.
3.3 OLM Noise Analysis Implementation The licensee provided as part of its LAR, the steps to implement the noise analysis technique to assess dynamic failure modes of pressure transmitters. The information provided in sections 3.3.1 through 3.3.6 of the LAR related to Diablo Canyon, is mapped to the steps found in section 11.3.3 Steps for Implementation of Noise Analysis Technique of AMS-TR-0720R2-A.
The NRC staff finds the provided mapping and information related to the noise analysis implementation consistent with the AMS OLM TR, AMS-TR-0720R2-A.
3.4 TS 1.1, Use and Application Definitions For Diablo Canyon, the TS definition for the term CHANNEL CALIBRATION is being revised to account for the approved OLM methodologies. The specific change allows transmitters that are included in the licensees OLM program to be excluded from the scope of instrumentation to be periodically calibrated within the frequency in the Surveillance Frequency Control Program (SFCP).
The NRC staff reviewed this proposed change considering the context of the OLM program.
This change is acceptable because the OLM processes would include an acceptable method for identifying performance issues as they occur and initiating corrective actions when pre-established OLM limits are exceeded. The corrective actions would also include performing instrument calibrations as necessary to restore instrument performance to within acceptable parameters. Data collected during OLM activities would be used to adjust OLM limits such that
poorly performing instruments would be calibrated at greater frequencies to address any potential impact on long term plant performance.
For Diablo Canyon, the TS definition for the terms, ENGINEERED SAFETY FEATURE (ESF)
RESPONSE TIME and REACTOR TRIP SYSTEM (RTS) RESPONSE TIME, are being revised to extend the current exclusion from periodic response time testing for instruments that are entered into the OLM program. The previous exclusion from response time testing had been based on the periodic channel calibration program, which will be replaced with the OLM program for those instruments that are included in the OLM scope.
The NRC staff finds these revised definitions to be acceptable because the OLM program will continue to monitor instrument performance and will be capable of detecting instrument degradation or failures that could affect response time performance. The previous definitions for these terms allowed exclusion from response time testing based on the fact that instrument failures that affect time response would be detectable during the periodic calibration tests and channel check activities. Since the OLM program will retain the capability of detecting and correcting instrument degraded performance or fault conditions, the NRC staff considers this method to be an acceptable and approved methodology to support continued exception of these instruments from time response testing.
3.5 New Diablo Canyon TS 5.5.21, Online Monitoring Program New TS 5.5.21 provides a description of the AMS based OLM program. The new TS stipulates that the OLM program must be implemented in accordance with the NRC-approved AMS OLM TR, AMS-TR-0720R2-A. Diablo Canyon TS 5.5.21 lists the key elements of the OLM program.
The NRC staff reviewed the TS description of the OLM program in the LAR and found that it is consistent with the program descriptions provided in the approved AMS OLM TR, AMS-TR-0720R2-A. To verify that the Diablo Canyon program would be implemented in accordance with the NRC approved TR, the NRC staff conducted an audit per audit plan dated February 25, 2025 (ML25051A270), and reviewed several Diablo Canyon specific reports that documented program implementation activities. These reports are described in the NRC staffs audit report dated August 19, 2025 (ML25230A241). The NRC staff audit confirmed that key elements including calculations of OLM limits, amenable transmitters to be included in the OLM program, backstop calculations, noise analysis implementation, maximum sampling rate calculations, OLM coverage of transmitter setpoints and range, drift monitoring, plant procedures for data retrieval, and analysis and capture of the OLM program would be implemented as described in AMS-TR-0720R2-A.
The NRC staff also reviewed Diablo Canyons responses to each of the Application Specific Action Items (ASAIs) that are contained in section 4.0 of the NRC SE for AMS OLM TR, AMS-TR-0720R2-A. These licensee responses are provided in section 3.4 of the LAR dated December 31, 2024. The NRC staff evaluated these ASAIs in section 3.6 of this SE. The NRC determined that all plant specific actions would be performed at an acceptable level and the Diablo Canyon OLM program would be implemented in conformance with the approved AMS OLM TR, AMS-TR-0720R2-A.
3.6 AMS TR-0720R2-A - Application Specific Action Items The NRC staff identified five ASAIs in the SE of the AMS OLM Program TR.
3.6.1 AMS-TR-0720R2-A ASAI 1 - Evaluation and Proposed Markup of Existing Plant Technical Specifications ASAI 1:
When preparing a license amendment request to adopt OLM methods for establishing calibration frequency, licensees should consider markups that provide clear requirements for accomplishing plant operations, engineering data analysis, and instrument channel maintenance. Such TS changes would need to include appropriate markups of the TS tables describing limiting conditions for operation and surveillance requirements, the technical basis for the changes, and the administrative programs section.
The licensee provided markups of the applicable TSs that provide clear requirements for accomplishing plant operations, engineering data analysis, and instrument channel maintenance for transmitters that are included in the OLM program. Markups of the TS Bases were also provided, which describe the technical basis for the OLM program. Therefore, the NRC staff finds the criteria of ASAI 1 are met.
3.6.2 AMS-TR-0720R2-A ASAI 2 - Identification of Calibration Error Source ASAI 2:
When determining whether an instrument can be included in the plant OLM program, the licensee shall evaluate calibration error source in order to account for the uncertainty due to multiple instruments used to support the transfer of transmitter signal data to the data collection system. Calibration errors identified through OLM should be attributed to the transmitter until testing can be performed on other support devices to correctly determine the source of calibration error and reallocate errors to these other loop components.
The NRC staff performed an audit of the Diablo Canyon OLM program reports to verify that calibration error sources were being factored into account for the uncertainty due to multiple instruments used to support the transfer of transmitter signal data to the data collection system.
The NRC staff found that the OLM program attributes calibration errors to the transmitter unless testing is subsequently performed to determine and reallocate calibration error to other instrument loop components. Therefore, the NRC staff finds the criteria of ASAI 2 are met.
3.6.3 AMS-TR-0720R2-A ASAI 3 - Response Time Test Elimination Basis ASAI 3:
If the plant has eliminated requirements for performing periodic RT [response time] testing of transmitters to be included in the OLM program, then the licensee shall perform an assessment of the basis for RT test elimination to determine if this basis will remain valid upon implementation of the OLM program and to
determine if the RT test elimination will need to be changed to credit the OLM program rather than the periodic calibration test program.
The transmitters that are being incorporated into the OLM program were excluded from response time testing. The licensee, therefore, performed an assessment of the basis for response time testing exclusions and determined that the OLM program will continue to support exclusion from response time testing, because the OLM methods will detect transmitter failures that would affect response time performance. The basis for this exclusion in TS 1.1 is evaluated in section 3.4 of this SE. Therefore, the NRC staff finds the criteria of ASAI 3 are met.
3.6.4 AMS-TR-0720R2-A ASAI 4 - Use of Calibration Surveillance Interval Backstop ASAI 4:
In its application for a license or license amendment to incorporate OLM methods for establishing calibration surveillance intervals, applicants or licensees should describe how they intend to apply backstop intervals as a means for mitigating the potential that a process groups could be experiencing undetected common mode drift characteristics.
The NRC staff performed an audit review of the backstop calculations performed for the Diablo Canyon transmitters being incorporated into the proposed OLM program and confirmed that these calculations were performed in a manner consistent with the processes outlined in the approved AMS OLM TR for determining maximum calibration intervals. Therefore, the NRC staff finds the criteria of ASAI 4 are met.
3.6.5 AMS-TR-0720R2-A ASAI 5 - Use of Criteria other than in AMS OLM TR for Establishing Transmitter Drift Flagging Limit ASAI 5:
In its application for a license or license amendment to incorporate OLM methods for establishing calibration surveillance intervals, applicants or licensees should describe whether they intend to adopt the criteria within the AMS OLM TR for flagging transmitter drift or whether they plan to use a different methodology for determining this limit.
The NRC staff determined that the Diablo Canyon proposed OLM program is consistent with the AMS OLM TR, AMS-TR-0720R2-A, and therefore, a different methodology is not being employed. Therefore, the NRC staff finds the criteria of ASAI 5 are met.
3.7 Technical Evaluation Summary The NRC staff finds that the licensees proposed implementation of the Diablo Canyon OLM program is consistent with the approved AMS OLM TR, AMS-TR -720R2-A. The NRC staff also finds the proposed revision to TS 1.1 and addition of TS 5.5.21 acceptable.
The NRC staff determined that implementation of the proposed OLM program for Diablo Canyon will continue to support establishment and maintenance of limiting safety system settings associated with the transmitters that are included in the program. These settings will continue to ensure that associated automatic protective actions will correct abnormal situations
before safety limits are exceeded. Implementation of the OLM program at Diablo Canyon would identify which protection system instrument channels require recalibration, which helps to ensure that the licensee would take appropriate actions if the licensee determines that an automatic safety system does not function as required. The surveillance requirements relating to test, calibration, and inspection of these transmitters will also continue to ensure that the adequate quality of systems and components is maintained. Therefore, the NRC staff finds that the requirements of 10 CFR 50.36(c)(1)(ii)(A), and 10 CFR 50.36(c)(3) will continue to be met.
Further, 10 CFR 50.36(c)(5) is met by addition of the new program to the licensees TSs.
Additionally, the NRC staff finds that the licensees implementation of the OLM Program in accordance with approved TR AMS-TR-720R2-A will continue to meet the requirements of the following principle design criteria for Diablo Canyon as documented in the Diablo Canyon UFSAR (as updated): pre-GDC 12, 14 and 15. The licensee is required to notify the NRC if an associated automatic safety system does not function as required.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the California State official was notified of the proposed issuance of the amendments on March 19, 2025. The State official had no comments.
5.0 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 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration published in the Federal Register on March 18, 2025 (90 FR 12572), and there has been no public comment on such finding.
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.
6.0 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: Gilberto Blas Rodriguez, NRR David Rahn, NRR Norbert Carte, NRR Samir Darbali, NRR Tarico Sweat, NRR Deanna Zhang, NRR Aaron Armstrong, NRR Date: August 21, 2025
- eConcurrence **by email OFFICE NRR/DORL/LPL4/PM*
NRR/DORL/LPL4/LA* NRR/DSS/STSB/BC** NRR/DEX/EICB/BC**
NAME SLee PBlechman SMehta (ARussell for) FSacko (JZhao for)
DATE 8/18/2025 8/20/2025 8/15/2025 8/14/2025 OFFICE NRR/DORL/LPL4/BC*
NRR/DORL/LPL4/PM*
NAME TNakanishi SLee DATE 8/21/2025 8/21/2025