NL-24-0064, Units 1 & 2 and Hatch Nuclear Plant - Units 1 & 2 License Amendment Request to Revise Technical Specification 1.1 and Add Online Monitoring Program to Technical Specification 5.5

Units 1 & 2 and Hatch Nuclear Plant - Units 1 & 2 License Amendment Request to Revise Technical Specification 1.1 and Add Online Monitoring Program to Technical Specification 5.5

ML24124A133
Person / Time
Site:	Farley, Hatch
Issue date:	05/03/2024
From:	Coleman J Southern Nuclear Operating Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
NL-24-0064
Download: ML24124A133 (1)

Text

., Southern Nuclear May 3, 2024 Docket Nos.: 50-321 50-348 50-366 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Regulatory Affairs Southern Nuclear Operating Company 3535 Colonnade Parkway Birmingham, AL 35243 205 992 5000 NL-24-0064 Farley Nuclear Plant - Units 1 &2 and Hatch Nuclear Plant - Units 1 &2 License Amendment Request to Revise Technical Specification 1.1 and Add Online Monitoring Program to Technical Specification 5.5 Ladies and Gentlemen:

Pursuant to the provisions Section 50.90 of Title 10 Code of Federal Regulations (CFR),

Southern Nuclear operating Company (SNC) hereby requests a license amendment to Farley Nuclear Plant (FNP) Unit 1 renewed operating license NPF-2 and Unit 2 renewed operating license NPF-8 and Hatch Nuclear Plant (HNP) Unit 1 renewed operating license DPR-57 and Unit 2 renewed operating license NPF-5. The proposed amendment revises Technical Specification (TS) 1.1, "Use and Application Definitions" and adds a new "Online Monitoring Program" to TS 5.5, "Programs and Manuals." SNC proposes to use online monitoring (OLM) methodology as the technical basis to switch from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on OLM results. The proposed change is based on the NRC-approved topical report AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters."

SNC requests approval of the proposed amendment six months following acceptance by the NRC. The proposed changes would be implemented within 60 days after issuance of the amendments.

In accordance with 10 CFR 50.91, a copy of this application is being provided to the designated Alabama and Georgia Officials.

This letter contains no regulatory commitments. If you have any questions, please contact Ryan Joyce at 205.992.6468.

U. S. Nuclear Regulatory Commission NL-24-0064 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 3rd day of May 2024.

Jamie M. Coleman Regulatory Affairs Director Southern Nuclear Operating Company JMC/kgl/cg

Enclosure:

Evaluation of the Proposed Change cc:

Regional Administrator, Region II NRR Project Manager - Farley, Hatch Senior Resident Inspector - Farley, Hatch Director, Alabama Office of Radiation Control Georgia Environmental Director Protection Division RType: Farley - CFA04.054; Hatch - CHA02.004

ENCLOSURE Evaluation of Proposed Change

1.

SUMMARY

DESCRIPTION

2.

DETAILED DESCRIPTION

2.1 Background

2.2 System Design and Operation 2.3 Reason for the Proposed Change 2.4 Description of Proposed Change

3. TECHNICAL EVALUATION 3.1 OLM Implementation Process Development 3.2 OLM Program Implementation 3.3 OLM Noise Analysis Implementation 3.4 Application Specific Action Items from AMS OLM TR

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination Analysis 4.4 Conclusions

5. ENVIRONMENTAL CONSIDERATION 6.REFERENCES ATTACHMENTS:

1.

Proposed Technical Specification Mark-ups -

FNP-1 &2

2.

Proposed Technical Specification Mark-ups -

HNP-1

3.

Proposed Technical Specification Mark-ups -

HNP-2

4.

Technical Specification Clean Typed -

FNP-1 &2

5.

Technical Specification Clean Typed -

HNP-1

6.

Technical Specification Clean Typed -

HNP-2

7.

Technical Specification Bases Mark-ups -

FNP-1&2 (Information only)

8.

Technical Specification Bases Mark-ups -

HNP-1 (Information only)

9.

Technical Specification Bases Mark-ups -

HNP-2 (Information only)

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Enclosure to NL-24-0064 Evaluation of Proposed Changes 1

SUMMARY

DESCRIPTION Pursuant to the provisions Section 50.90 of Title 10 Code of Federal Regulations (CFR),

Southern Nuclear Operating Company (SNC) hereby requests a license amendment to the Joseph M Farley Nuclear Generating Plant Units 1 and 2 (FNP) operating licenses NPF-2 and NPF-8, respectively, and Edwin I. Hatch Nuclear Power Plant Units 1 and 2 (HNP-1 and HNP-2) operating licenses DPR-57 and NPF-5, respectively. The proposed amendment revises Definitions and adds a new "Online Monitoring Program." SNC proposes to use online monitoring (OLM) methodology as the technical basis to switch from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on OLM results.

2 DETAILED DESCRIPTION

2.1 Background

OLM technologies have been developed and validated for condition monitoring applications in a variety of process and power industries. This application of OLM is used to optimize maintenance of instrumentation and control (l&C) systems including online drift monitoring and assessment of dynamic failure modes of transmitters. Analysis and Measurement Services (AMS) Topical Report (TR) AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters" (References 1 and 2) focused on the application of OLM for monitoring drift of pressure, level, and flow transmitters in nuclear power plants. The TR addressed the following topics:

Advances in OLM implementation technology to extend transmitter calibration intervals Experience with OLM implementation in nuclear facilities Comparison between OLM results and manual calibrations Transmitter failure modes that can be detected by OLM Related regulatory requirements and industry standards and guidelines Procedures for implementation of OLM methodology Changes that must be made to existing technical specifications to adopt OLM AMS-TR-0720R2-A provided the NRC with the information needed to approve the AMS OLM methodology for implementation in nuclear power plants. The TR is intended to be used by licensees to support plant-specific technical specification changes to switch from time-based calibration frequency of pressure, level, and flow transmitters to a condition-based calibration frequency based on OLM results and to develop procedures to assess dynamic failure modes of pressure sensing systems using the noise analysis technique.

The NRC staff determined that the methodology outlined in the AMS OLM TR for applying OLM techniques to pressure, level, and flow transmitters can be used to provide reasonable assurance that required TS instrument calibration requirements for transmitters will be maintained. This determination was based on the NRC staff finding that OLM techniques: a) are effective at identifying instrument calibration drift during plant operation, b) provide an acceptable means of identifying when manual transmitter calibration using traditional calibration methods are needed, and c) will maintain an acceptable level of performance that is traceable to calibration prime standards.

The NRC staff found that implementation of an OLM program in accordance with the approved AMS OLM TR provides an acceptable alternative to periodic manual calibration surveillance E-2

Enclosure to NL-24-0064 Evaluation of Proposed Changes requirements upon implementation of the application-specific action items (ASAI) in Section 4.0 of its safety evaluation. The ASAls are addressed in Section 3.4 below.

2.2 System Design and Operation 2.2.1 Farley Nuclear Plant The transmitters to be included in the Online Monitoring Program provide input to the Reactor Trip System (RTS), Engineered Safety Feature Actuation System (ESFAS), Post Accident Monitoring (PAM), and Remote Shutdown System.

The RTS initiates a unit shutdown, based on the values of selected unit parameters, to protect against violating the core fuel design limits and Reactor Coolant System (RCS) pressure boundary during anticipated operational occurrences and to assist the Engineered Safety Features Systems in mitigating accidents. The RTS and related instrumentation are identified in TS Table 3.3.1-1.

The ESFAS is used to actuate engineered safety features and related systems initiate necessary safety systems, based on the values of selected unit parameters, to protect against violating core design limits and the RCS pressure boundary, and to mitigate accidents. The systems used to actuate engineered safety features and related instrumentation are identified in TS Table 3.3.2-1.

The primary purpose of the PAM instrumentation is to display unit variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for Design Basis Accidents. The PAM instrumentation is identified in TS Table 3.3.3-1.

The Remote Shutdown System provides the operator with sufficient instrumentation and controls to place and maintain the unit in a safe shutdown condition from a location other than the control room. This capability is necessary to protect against the possibility that the control room becomes inaccessible. The Remote Shutdown System instrumentation is identified in TS 3.3.4.

The RTS, ESFAS, PAM, and Remote Shutdown System transmitters were evaluated in accordance with the methodology in AMS-TR-0720R2-A. The transmitters to be included in the OLM program and the bases for their selection can be found in AMS report FNP2301 RO, "OLM Amenable Transmitters Report for Farley Units 1 & 2" (Reference 3).

Switching from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency will not create any physical changes to the plants. The changes will not impact how the plants operate. SNC will use condition-based frequency to determine when transmitter calibrations are needed instead of performing calibrations based on a calendar frequency. Existing calibration methods will be used when it is determined that transmitter calibration is needed.

2.2.2 Hatch Nuclear Plant Unit 1 The transmitters to be included in the Online Monitoring Program provide input to the Reactor Protection System (RPS), BWR Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) System, Emergency Core Cooling System (ECCS) Instrumentation and Primary Containment Isolation Instrumentation used to actuate engineered safety features, PAM, and the Remote Shutdown System.

The RPS initiates a unit shutdown, based on the values of selected unit parameters, to protect against violating the core fuel design limits and RCS pressure boundary during anticipated E-3

Enclosure to NL-24-0064 Evaluation of Proposed Changes operational occurrences and to assist the Engineered Safety Features Systems in mitigating accidents. The BWR ATWS-RPT System initiates a RPT, adding negative reactivity, following events in which a scram does not (but should) occur, to lessen the effects of an ATWS event.

These systems and related instrumentation are identified in TS Table 3.3.1.1-1 and TS 3.3.4.2.

The systems used to actuate engineered safety features and related systems initiate necessary safety systems, based on the values of selected unit parameters, to protect against violating core design limits and the RCS pressure boundary, and to mitigate accidents. The systems used to actuate engineered safety features and related instrumentation are identified in TS Table 3.3.5.1-1 and Table 3.3.6.1-1.

The primary purpose of the PAM instrumentation is to display unit variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for Design Basis Accidents. The PAM instrumentation is identified in TS Table 3.3.3.1-

1.

The Remote Shutdown System provides the operator with sufficient instrumentation and controls to place and maintain the unit in a safe shutdown condition from a location other than the control room. This capability is necessary to protect against the possibility that the control room becomes inaccessible. The Remote Shutdown System instrumentation is addressed in TS 3.3.3.2.

The RPS, systems used to actuate engineered safety features, PAM, and Remote Shutdown System transmitters were evaluated in accordance with the methodology in AMS-TR-0720R2-A.

The transmitters to be included in the OLM program and the bases for their selection can be found in AMS report HAT2301 RO, "OLM Amenable Transmitters Report for Hatch Units 1 & 2" (Reference 4 ).

Switching from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency will not create any physical changes to the plants. The changes will not impact how the plants operate. SNC will use condition-based frequency to determine when transmitter calibrations are needed instead of performing calibrations based on a calendar frequency. Existing calibration methods will be used when it is determined that transmitter calibration is needed.

2.2.3 Hatch Nuclear Plant Unit 2 The transmitters to be included in the Online Monitoring Program provide input to the Reactor Protection System (RPS), BWR ATWS-RPT System, ECCS Instrumentation and Primary Containment Isolation Instrumentation used to actuate engineered safety features, PAM, and the Remote Shutdown System.

The RPS initiates a unit shutdown, based on the values of selected unit parameters, to protect against violating the core fuel design limits and RCS pressure boundary during anticipated operational occurrences and to assist the Engineered Safety Features Systems in mitigating accidents. The BWR ATWS-RPT System initiates a RPT, adding negative reactivity, following events in which a scram does not (but should) occur, to lessen the effects of an ATWS event.

These systems and related instrumentation are identified in TS Table 3.3.1.1-1 and TS 3.3.4.2.

The systems used to actuate engineered safety features and related systems initiate necessary safety systems, based on the values of selected unit parameters, to protect against violating core design limits and the RCS pressure boundary, and to mitigate accidents. The systems E-4

Enclosure to NL-24-0064 Evaluation of Proposed Changes used to actuate engineered safety features and related instrumentation are identified in TS Table 3.3.5.1-1 and Table 3.3.6.1-1.

The primary purpose of the PAM instrumentation is to display unit variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for Design Basis Accidents. The PAM instrumentation is identified in TS Table 3.3.3.1-

1.

The Remote Shutdown System provides the operator with sufficient instrumentation and controls to place and maintain the unit in a safe shutdown condition from a location other than the control room. This capability is necessary to protect against the possibility that the control room becomes inaccessible. The Remote Shutdown System instrumentation is addressed in TS 3.3.3.2.

The RPS, systems used to actuate engineered safety features, PAM, and Remote Shutdown System transmitters were evaluated in accordance with the methodology in AMS-TR-0720R2-A.

The transmitters to be included in the OLM program and the bases for their selection can be found in the following AMS report HAT2301 RO, "OLM Amenable Transmitters Report for Hatch Units 1 & 2" (Reference 4).

Switching from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency will not create any physical changes to the plants. The changes will not impact how the plants operate. SNC will use condition-based frequency to determine when transmitter calibrations are needed instead of performing calibrations based on a calendar frequency. Existing calibration methods will be used when it is determined that transmitter calibration is needed.

2.3 Reason for the Proposed Change SNC is proposing to use the NRC-approved OLM methodology described in AMS-TR-0720R2-A. The use of the NRC-approved OLM methodology ensures that plant safety is maintained by demonstrating that transmitters are functioning correctly. The OLM methodology encompasses environmental and process conditions in the assessment of transmitter calibration.

The use of condition-based monitoring for transmitter calibration provides additional safety benefits, as described in AMS-TR-0720R2-A. The use of OLM will result in elimination of unnecessary transmitter calibration and associated opportunities for human errors. Elimination of unnecessary calibrations will also reduce calibration-induced damage to transmitters and other plant equipment. The use of OLM provides for timely detection of out-of-calibration transmitters. It eliminates occupational exposure or human error opportunities related to calibration activities that were unnecessary. Experience has shown that human errors during calibration of transmitters that did not require recalibration have resulted in additional repairs to correct the mistakes.

2.4 Description of the Proposed Change 2.4.1 Farley Nuclear Plant SNC proposes to change TS 1.1 "Use and Application Definitions" definition of CHANNEL CALIBRATION.

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Enclosure to NL-24-0064 Evaluation of Proposed Changes Current TS 1.1 CHANNEL CALIBRATION -A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel so that it responds within the required range and accuracy to known input. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor, alarm, interlock, and trip functions. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION shall include and inplace cross calibration that compares the other sensing elements with the recently installed sensing element. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

Proposed TS 1.1 CHANNEL CALIBRATION -A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel so that it responds within the required range and accuracy to known input. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor (excluding transmitters in the Online Monitoring Program), alarm, interlock, and trip functions. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION shall include and inplace cross calibration that compares the other sensing elements with the recently installed sensing element. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

SNC proposes to change TS 1.1 "Use and Application Definitions" definition of ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME.

Current TS 1.1 ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME -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 the 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.

Proposed TS 1.1 ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME - 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 E-6

Enclosure to NL-24-0064 Evaluation of Proposed Changes 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 the methodology for verification have been previously reviewed and approved by the NRC (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.

SNC proposes to change TS 1.1 "Use and Application Definitions" definition of REACTOR TRIP SYSTEM (RTS) RESPONSE TIME.

Current TS 1.1 REACTOR TRIP SYSTEM (RTS) RESPONSE TIME -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 the 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.

Proposed TS 1.1 REACTOR TRIP SYSTEM (RTS) RESPONSE TIME -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 the methodology for verification have been previously reviewed and approved by the NRC (including transmitters in the Online Monitoring Program),

or the components have been evaluated in accordance with an NRC approved methodology.

SNC proposes to add a new Online Monitoring TS 5.5.21:

Online Monitoring Program This program provides controls to determine the need for calibration of pressure, level, and flow transmitters using condition monitoring based on drift analysis.

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.

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.

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Enclosure to NL-24-0064 Evaluation of Proposed Changes

c.

Performance of calibration checks for transmitters 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.

2.4.2 Hatch Nuclear Plant Unit 1 SNC proposes to change TS 1.1 "Use and Application Definitions" definition of CHANNEL CALIBRATION.

Current TS 1.1 CHANNEL CALIBRATION -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 the entire channel, including the required sensor, alarm, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace 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, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

Proposed TS 1.1 CHANNEL CALIBRATION -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 the entire channel, including the required sensor (excluding transmitters in the Online Monitoring Program), alarm, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace 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, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

SNC proposes to change TS 1.1 "Use and Application Definitions" definition of REACTOR PROTECTION SYSTEM (RPS) RESPONSE TIME.

Current TS 1.1 The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. 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.

Proposed TS 1.1 The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. 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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

SNC proposes to add a new Online Monitoring TS 5.5.17:

E-8

Enclosure to NL-24-0064 Evaluation of Proposed Changes Online Monitoring Program This program provides controls to determine the need for calibration of pressure, level, and flow transmitters using condition monitoring based on drift analysis.

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.

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 transmitters 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.

2.4.3 Hatch Nuclear Plant Unit 2 SNC proposes to change TS 1.1 "Use and Application Definitions" definition of CHANNEL CALIBRATION.

Current TS 1.1 CHANNEL CALIBRATION -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 the entire channel, including the required sensor, alarm, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace 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, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

Proposed TS 1.1 CHANNEL CALIBRATION -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 the entire channel, including the required sensor (excluding transmitters in the Online Monitoring Program), alarm, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST. Calibration of instrument channels with resistance temperature detector E-9

Enclosure to NL-24-0064 Evaluation of Proposed Changes (RTD) or thermocouple sensors may consist of an inplace 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, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

SNC proposes to change TS 1.1 "Use and Application Definitions" definition of EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME.

Current TS 1.1 The ECCS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS initiation setpoint at the channel sensor until the ECCS 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.

Proposed TS 1.1 The ECCS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS initiation setpoint at the channel sensor until the ECCS 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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

SNC proposes to change TS 1.1 "Use and Application Definitions" definition of ISOLATION SYSTEM RESPONSE TIME.

Current TS 1.1 The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence 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.

Proposed TS 1.1 The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence 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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

SNC proposes to change TS 1.1 "Use and Application Definitions" definition of REACTOR PROTECTION SYSTEM (RPS) RESPONSE TIME.

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Enclosure to NL-24-0064 Evaluation of Proposed Changes Current TS 1.1 The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. 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.

Proposed TS 1.1 The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. 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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

SNC proposes to add a new Online Monitoring TS 5.5.17:

Online Monitoring Program This program provides controls to determine the need for calibration of pressure, level, and flow transmitters using condition monitoring based on drift analysis.

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.

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 transmitters 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.

The proposed TS changes are an adaptation from the illustrative changes presented in AMS-TR-0720R2-A that simplify the required plant-specific changes. The proposed Definition changes eliminated the need to modify the Channel Calibration and Response Time Surveillance Requirements. The proposed Online Monitoring Program description was reorganized to better align with the OLM implementation activities.

3 TECHNICAL EVALUATION E-11

Enclosure to NL-24-0064 Evaluation of Proposed Changes 3.1 OLM Implementation Process Development This section describes the steps that were performed to implement the OLM program for FNP, HNP-1, and HNP-2 by following the steps identified in AMS-TR-0720R2-A Section 11.1.1. This work is documented in the AMS reports on OLM Amenable Transmitters (References 3 and 4) and OLM Analysis Methods and Limits (References 5 and 6).

The AMS reports on OLM Amenable Transmitters address steps 1-6, from AMS-TR-0720R2-A Section 11.1.1. These steps were designed to arrive at a list of transmitters that can be included in an OLM program and determine how to obtain OLM data. The RTS, RPS, systems used to actuate engineered safety features, PAM, and Remote Shutdown System transmitters to be included in the OLM program and the bases for their selection can be found in the AMS reports on OLM Amenable Transmitters (References 3 and 4 ).

3.1.1 Determine if Transmitters are Amenable to OLM AMS-TR-0720R2-A Chapter 12 includes Table 12.4 that lists the nuclear grade transmitter models that are amenable to OLM. Any transmitter model that is not listed in this table should only be added to the OLM program if it can be shown by similarity analysis that its failure modes are the same as the listed transmitter models or otherwise detectable by OLM.

3.1.2 List Transmitters in Each Redundant Group This step establishes how to group the transmitters and evaluates the redundancy of each group.

3.1.3 Determine if OLM Data Covers Applicable Setpoints This step evaluates the OLM data for each group to determine if it covers applicable setpoints.

Additional details are described in AMS-TR-0720R2-A Chapter 14.

3.1.4 Calculate Backstops A backstop, as described in AMS-TR-0720R2-A Chapter 13, must be established for each group of redundant transmitters amenable to OLM as a defense against common mode drift.

The backstop identifies the maximum period between calibrations without calibrating at least one transmitter in a redundant group.

3.1.5 Establish Method of Data Acquisition OLM data is normally available in the plant computer or an associated data historian. If data is not available from the plant computer or historian, a custom data acquisition system including hardware and software must be employed to acquire the data.

3.1.6 Specify Data Collection Duration and Sampling Rate OLM data must be collected during startup, normal operation, and shutdown periods at the highest sampling rate by which the plant computer takes data. AMS-TR-0720R2-A Chapter 15 describes a process to determine the minimum sampling rate for OLM data acquisition to monitor for transmitter drift. AMS-TR-0720R2-A Chapter 8 describes a process to help determine the optimal sampling rate and minimum duration of OLM data collection.

AMS reports on OLM Analysis Methods and Limits (References 5 and 6) address steps 7-8, from AMS-TR-0720R2-A Section 11.1.1 These steps address the calculation of the OLM limits and establish the methods of OLM data analysis.

3.1.7 Identify Data Analysis Methods OLM implementations must employ both simple averaging and parity space methods for data analysis as described in AMS-TR-0720R2-A Chapter 6.

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Enclosure to NL-24-0064 Evaluation of Proposed Changes 3.1.8 Establish OLM Limits OLM limits must be established as described in AMS-TR-0720R2-A Chapter 7 for each group of redundant transmitters. Calculation of OLM limits must be based on combining uncertainties of components of each instrument channel from the transmitter in the field to the OLM data storage.

The AMS report on OLM Analysis Methods and Limits provides the OLM Limit calculations for the transmitters that are amenable to OLM at FNP, HNP-1, and HNP-2.

3.2 OLM Program Implementation This section summarizes the steps that must be followed to implement the OLM program for transmitter drift monitoring at FNP, HNP-1, and HNP-2 in accordance with AMS-TR-0720R2-A.

The steps described in this section are repeated at each operating cycle at FNP, HNP-1, and HNP-2 to identify the transmitters that should be scheduled for a calibration check using data from periods of startup, normal operation, and shutdown. Additional details regarding the OLM Program Implementation discussed in this section are contained in the AMS reports on OLM Drift Monitoring Implementation (References 7 and 8).

AMS-TR-0720R2-A Section 11.1.2 identifies eleven steps that must be followed each operating cycle to identify the transmitters that should be scheduled for a calibration check at the ensuing outage. Table 1 provides a mapping between AMS-TR-0720R2-A Section 11.1.2 and the LAR section where the item is addressed. Implementation of these steps is performed using the AMS Bridge and the AMS Calibration Reduction System (CRS) software programs that were developed by AMS under their 10 CFR Part 50 Appendix B software Quality Assurance (QA) program.

Table 1: Mapping to AMS-TR-0720R2-A Section 11.1.2 Step Number in Section LAR Item Step 11.1.2 of Section AMS-TR-0720R2-A 1

Retrieve OLM Data 9

3.2.1 2

Perform Data Qualification 10 3.2.2 3

Select Appropriate Region of Any Transient Data 11 3.2.3 4

Perform Data Analysis 12 3.2.4 5

Plot the Average Deviation for Each Transmitter 13 3.2.5 6

Produce a Table for Each Group That Combines All 14 3.2.6 Results 7

Determine OLM Results for Each Transmitter 15 3.2.7 8

Address Uncertainties in the Unexercised Portion of 16 3.2.8 Transmitter Range 9

Select Transmitters to Be Checked for Calibration as a 17 3.2.9 Backstop 10 Perform Dynamic Failure Mode Assessment 18 3.2.10 E-13

Enclosure to NL-24-0064 Evaluation of Proposed Changes Item I

Step Produce a Report of Transmitters Scheduled for Calibration Check 3.2.1 Retrieve OLM Data I

Step Number in Section 11.1.2 of AMS-TR-0720R2-A 19 LAR Section The first step in performing transmitter drift monitoring is to retrieve the OLM data. OLM data must be retrieved during periods of startup, normal operation, and shutdown. The method of data acquisition, data collection duration, sampling rate, and list of sensors whose data will be retrieved have been established as described in Section 3.1 of this document. The OLM data for FNP, HNP-1, and HNP-2 will be retrieved using the AMS Bridge software which will retrieve data from the Southern Nuclear Maintenance and Diagnostic (M&D) center historian and produce binary data files that are compatible with the AMS Calibration Reduction System (CRS) software or as a text files from the SNC data historian or other data sources at each plant site, as applicable. AMS procedure OLM2201, "Procedure for Online Monitoring Data Retrieval," has been developed for performing the data retrieval using the AMS Bridge software (Reference 9).

3.2.2 Perform Data Qualification OLM data retrieved from plant historians sometimes contains anomalies such as spikes, missing data, stuck data, and saturated data. The portion of data containing these anomalies should be excluded, filtered, and/or cleaned prior to analysis. The AMS CRS software provides functionality for these tasks and will be used to perform data qualification. AMS procedure OLM2202, "Procedure for Performing Online Monitoring Data Qualification and Analysis," has been developed for performing data qualification and analysis using the AMS CRS software (Reference 10).

3.2.3 Select Appropriate Region of Any Transient Data The AMS CRS software provides means to select the regions of transient data as described in Step 11 of Section 11.1.2 of AMS-TR-0720R2-A and will be used to perform these selections.

This activity is part of OLM data analysis and is addressed in the data qualification and analysis procedure.

3.2.4 Perform Data Analysis Several tasks that must be performed in OLM data analysis for startup, normal operation, and shutdown data including:

1. Calculate the process estimate,

2. Calculate the deviation of each transmitter from the process estimate and plot the

outcome,

3. Partition the deviation data into region(s) by percent of span,

4. Calculate and plot the average deviation for each region versus percent of span,

5. Select appropriate process estimation techniques, filtering parameters, and remove any

outliers,

6. Determine if average deviations exceed OLM limits for any region, and

7. Review, document, and store the details and results of analysis.

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Enclosure to NL-24-0064 Evaluation of Proposed Changes The AMS CRS software provides functionality for performing these tasks and will be used to perform OLM data analysis. Detailed steps for performing OLM data analysis are provided in the data qualification and analysis procedure.

3.2.5 Plot the Average Deviation for Each Transmitter The AMS CRS software provides functionality for plotting the average deviation for each transmitter as described in Step 13 of Section 11.1.2 of AMS-TR-0720R2-A and will be used to perform this task. This activity is part of OLM data analysis and is addressed in detail in the data qualification and analysis procedure.

3.2.6 Produce a Table for Each Group That Combines All Results The AMS CRS software provides functionality for producing a table for each group of redundant transmitters that combines all results as described in Step 13 of Section 11.1.2 of AMS-TR-0720R2-A and will be used to perform this task. This activity is part of OLM data analysis and is addressed in detail in the data qualification and analysis procedure.

3.2.7 Determine OLM Results for Each Transmitter OLM results must be produced by the OLM analyst upon completion of data analysis for a complete operating cycle. The AMS CRS software provides functionality for producing these results as described in Step 15 of Section 11.1.2 of AMS-TR-0720R2-A and will be used to perform this task. This activity is part of OLM data analysis and is addressed in detail in the data qualification and analysis procedure.

3.2.8 Address Uncertainties in the Unexercised Portion of Transmitter Range The AMS CRS software provides functionality for addressing uncertainties in the unexercised portion of the transmitter ranged as described in Step 13 of Section 11.1.2 of AMS-TR-0720R2-A and will be used to perform this task. This activity is part of OLM data analysis and is addressed in detail in the data qualification and analysis procedure.

3.2.9 Select Transmitters to Be Checked for Calibration as a Backstop The AMS procedure OLM2202 is also used for maintaining the backstops for OLM. It provides detailed steps for selecting transmitters to be checked for calibration as a backstop as described in Step 17 of Section 11.1.2 of AMS-TR-0720R2-A.

3.2.10 Perform Dynamic Failure Mode Assessment As described in Step 18 of Section 11.1.2 of AMS-TR-0720R2-A, dynamic failure mode assessment must be performed using the noise analysis technique to cover dynamic failures that are not detectable by the OLM process for transmitter drift monitoring. Details on how this will be addressed for FNP, HNP-1, and HNP-2 are described in LAR Section 3.3.

3.2.11 Produce a Report of Transmitters Scheduled for Calibration Check The results of OLM analysis must be compiled in a report and independently reviewed. The transmitters that have been flagged must be scheduled for a calibration check at the next opportunity. The AMS CRS software provides functionality for producing this report and will be used to perform this task. This activity is part of OLM data analysis and is addressed in detail in the data qualification and analysis procedure.

3.3 OLM Noise Analysis Implementation Some licensees have extended or eliminated transmitter response time testing requirements with NRC approval based, in part, on the performance of manual calibrations. Manual calibrations will not be performed except on transmitters that are flagged by OLM. The noise E-15

Enclosure to NL-24-0064 Evaluation of Proposed Changes analysis methodology is provided in this document to enable licensees to assess the dynamic failure modes of transmitters that are not covered by the OLM process for transmitter drift monitoring.

This section summarizes the steps that must be followed to implement the noise analysis technique for transmitter dynamic failure mode assessment at FNP, HNP-1, and HNP-2 in accordance with AMS-TR-0720R2-A. Additional details regarding the implementation of the noise analysis technique discussed in this section are provided in the AMS reports on Noise Analysis Programs (References 11 and 12).

As described in Section 11.3.3 of AMS-TR-0720R2-A, six steps must be followed to assess dynamic failure modes of pressure transmitters. Table 2 provides a mapping of the six steps in Section 11.3.3 of AMS-TR-0720R2-A and the section where they are addressed in this document. Implementation of these steps is performed using qualified noise data acquisition equipment and software programs that were developed by AMS under their 10 CFR Part 50 Appendix B software Quality Assurance (QA) program.

For FNP, HNP-1, and HNP-2, the transmitters with response time requirements have been identified in AMS reports on OLM Amenable Transmitters (References 3 and 4).

Table 2: Mapping to AMS-TR-0720R2-A Section 11.3.3 Item Step Step Number in Section 11.3.3 of LAR AMS-TR-0720R2-A Section 1

Select Qualified Noise Data Acquisition 1

3.3.1 Equipment 2

Connect Noise Data Acquisition 2

3.3.2 Equipment to Plant Signals 3

Collect and Store Data for Subsequent 3

3.3.3 Analysis 4

Screen Data for Artifacts and Anomalies 4

3.3.4 5

Perform Data Analysis 5

3.3.5 6

Review and Document Results 6

3.3.6 3.3.1 Select Qualified Noise Data Acquisition Equipment The first step in performing noise analysis is to select qualified noise data acquisition equipment. This equipment must have a valid calibration traceable to the National Institute of Standards and Technology and meet a set of performance criteria detailed Step 1 of Section 11.3.3 of AMS-TR-0720R2-A. The equipment used to acquire data at FNP, HNP-1, HNP-2 will be the AMS OLM data acquisition system which is comprised of hardware and software that has been developed and tested using AMS 10 CFR Part 50 Appendix B hardware and software QA program.

3.3.2 Connect Noise Data Acquisition Equipment to Plant Signals AMS Procedure NPS1501, "Procedure for Noise Data Collection from Plant Sensors," is used for the connection of the noise data acquisition equipment for performing noise analysis testing E-16

Enclosure to NL-24-0064 Evaluation of Proposed Changes (Reference 13). This procedure identifies the locations for connection to process signals as well as the qualified personnel who may connect the data acquisition system at these locations. The noise data acquisition system should be connected to as many transmitters as allowed by the number of data acquisition channels and the plant procedures. Multiple transmitters (e.g., up to

32) can be tested simultaneously to reduce the test time. Each data acquisition channel must be connected to the transmitter current loop as shown in Section 11.3.3 of AMS-TR-0720R2-A.

3.3.3 Collect and Store Data for Subsequent Analysis The noise data should be collected during normal plant operation at full temperature, pressure, and flow and analyzed in real time or stored to be analyzed later. However, noise data taken at other conditions is acceptable as long as there is enough process fluctuation with sufficient amplitude and frequency content to drive the transmitters to reveal their dynamic characteristics.

Noise data collection will be performed using AMS OLM Data Acquisition software which has been developed and tested using AMS software V&V program which conforms to 10 CFR Part 50 Appendix B. The use of this software for noise data acquisition is addressed in the AMS procedure for performing noise analysis testing (Reference 13).

3.3.4 Screen Data for Artifacts and Anomalies Noise data may contain anomalies that must be excluded, filtered, and/or cleaned prior to data analysis. AMS Procedure NAR2201, "Procedure for Performing Dynamic Failure Mode Assessment Using Noise Analysis," is used for performing noise analysis data analysis (Reference 14) and will be performed using AMS noise analysis software.

3.3.5 Perform Data Analysis Noise data analysis will be performed as described in Section 11.3.3 Step 5 in AMS-TR-0720R2-A using AMS noise analysis software. General data analysis steps for the analyst as well as detailed steps for performing noise data analysis are also provided in the AMS procedure for performing noise analysis data analysis (Reference 14 ).

3.3.6 Review and Document Results Results of noise data analysis will be reviewed and approved by qualified personnel and documented in a report. This process is detailed in the AMS procedure for performing noise analysis data analysis (Reference 14 ).

3.4 Application Specific Action Items from AMS OLM TR The NRC approval of the AMS OLM TR required implementation of the ASAls in Section 4.0 of its safety evaluation. Five ASAls were identified, and each is addressed below.

ASAI 1 - Evaluation and Proposed Mark-up of Existing Plant Technical Specifications 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.

Response to ASAI 1: The proposed changes to the FNP, HNP-1, and HNP-2 Technical Specifications are identified in Section 2.4 and shown in Attachments 1 and 2. The proposed changes modify applicable Definitions and add a new program for OLM in the Administrative E-17

Enclosure to NL-24-0064 Evaluation of Proposed Changes Controls. No changes to the Technical Specification tables describing Limiting Conditions for Operation or Surveillance Requirements were necessary.

ASAI 2 - Identification of Calibration Error Source 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.

Response to ASAI 2: Calibration error is evaluated as part of the calculation of OLM limits as described in Section 3.1.8. The calculation of OLM limits is based on combining uncertainties of components of each instrument channel from the transmitter in the field to the OLM data storage. The OLM data assessment methods described in Section 3.2.7 include guidance to consider calibration errors identified through OLM as coming from 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.

ASAI 3 - Response Time Test Elimination Basis If the plant has eliminated requirements for performing periodic RT 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.

Response to ASAI 3: FNP, HNP-1, and HNP-2 previously eliminated requirements for performing periodic response time testing based on the periodic calibration of transmitters that are proposed to be included in the OLM program. FNP, HNP-1, and HNP-2 propose to change the basis for response time test elimination to the methodology described in Section 3.3, which is based on the noise analysis methodology described in Section 11.3 of the AMS OLM TR.

ASAI 4 - Use of Calibration Surveillance Interval Backstop 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 group could be experiencing undetected common mode drift characteristics.

Response to ASAI 4: The SNC OLM programs for FNP, HNP-1, and HNP-2 adopt the calibration surveillance interval backstop methods described in Section 3.2.9, which is based on the backstop methodology described in Section 13 of the AMS OLM TR.

The Updated Final Safety Analysis Reports (UFSAR) for FNP, HNP-1, and HNP-2 will be modified to add the use of AMS-TR-0720R2-A.

Farley Nuclear Plant UFSAR Sections 7.1.2.7, 7.2.2.2.4, and 7.2.3.1 will be modified to add the following discussion:

Online monitoring using an NRC-approved methodology is used to switch from time-based calibration frequency of pressure, level, and flow transmitters to a condition-based calibration frequency based on online monitoring results. The E-18

Enclosure to NL-24-0064 Evaluation of Proposed Changes online monitoring methodology is also used to assess dynamic failure modes of pressure-type sensing systems using the noise analysis technique to support the use allocations for transmitter response times in lieu of response time tests. The transmitters included in the Online Monitoring Program are listed in Table 7.1-2.

UFSAR Sections 7.3.2.5 and 7.4.1 will be modified to add the following discussion:

Online monitoring using an NRC-approved methodology is used to switch from time-based calibration frequency of pressure, level, and flow transmitters to a condition-based calibration frequency based on online monitoring results. The transmitters included in the Online Monitoring Program are listed in Table 7.1-2.

UFSAR Sections 7.2.1.2, 7.3.1.2, and 7.3.2.7 and Tables 7.2-5 and 7.3-16 will be modified to add the following discussion:

Online monitoring using an NRC-approved methodology is used to assess dynamic failure modes of pressure-type sensing systems using the noise analysis technique to support the use allocations for transmitter response times in lieu of response time tests. The transmitters included in the Online Monitoring Program are listed in Table 7.1-2.

The following reference will be added to UFSAR Sections 7.1, 7.2, and 7.3:

Hashemian, H. M., Shumaker, B. D., and Morton, G. W., "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters," AMS-TR-0720R2-A, August 2021 Hatch Nuclear Plant Unit 1 UFSAR Sections 7.1.1, 7.2.5, 7.3.6, 7.4.5, 7.8.7, and 7.18.2.9:

Online monitoring using an NRC-approved methodology is used to switch from time-based calibration frequency of pressure, level, and flow transmitters to a condition-based calibration frequency based on online monitoring results. The online monitoring methodology is also used to assess dynamic failure modes of pressure-type sensing systems using the noise analysis technique to support the use allocations for transmitter response times in lieu of response time tests. The transmitters included in the Online Monitoring Program are listed in Table 7.1-3.

The following reference will be added to UFSAR Sections 7.2, 7.3, 7.4, 7.8, and 7.18:

Hashemian, H. M., Shumaker, B. D., and Morton, G. W., "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters," AMS-TR-0720R2-A, August 2021 Hatch Nuclear Plant Unit 2 UFSAR Sections 7.1.2.7, 7.2.3.3, 7.3.1.3.4, 7.3.2.3, and 7.4.1.3.4:

Online monitoring using an NRC-approved methodology is used to switch from time-based calibration frequency of pressure, level, and flow transmitters to a condition-based calibration frequency based on online monitoring results. The online monitoring methodology is also used to assess dynamic failure modes of pressure-type sensing systems using the noise analysis technique to support the use allocations for transmitter response times in lieu of response time tests. The transmitters included in the Online Monitoring Program are listed in Table 7.1-4.

UFSAR Sections 7.2.2.8, 7.3.1.3.1, and 7.3.2.3:

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Enclosure to NL-24-0064 Evaluation of Proposed Changes Alternately, the online monitoring methodology is used to assess dynamic failure modes of pressure-type sensing systems using the noise analysis technique to support the use allocations for transmitter response times in lieu of response time tests. The transmitters included in the Online Monitoring Program are listed in Table7.1-4.

The following reference will be added to UFSAR Sections 7.1, 7.2, 7.3, and 7.4:

Hashemian, H. M., Shumaker, B. D., and Morton, G. W., "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters," AMS-TR-0720R2-A, August 2021 ASAI 5 - Use of Criteria other than in AMS OLM TR for Establishing Transmitter Drift Flagging Limit 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.

Response to ASAI 5: The SNC OLM program for FNP, HNP-1, HNP-2 adopt the two averaging techniques (i.e., simple average and parity space) described in Section 6 of the AMS OLM TR for flagging transmitter drift.

4 REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.1.1 1 O CFR 50.36 Technical Specifications 10 CFR 50.36 Technical Specifications. Part (3) of this regulation sets the governing requirements for the inclusion of Surveillance Requirements in the Technical Specifications included in the Operating License for a commercial nuclear power plant.

(3) Surveillance requirements. 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.

SNC proposes to use the AMS OLM methodology for FNP, HNP-1, and HNP-2 as the technical basis to support plant-specific Technical Specification changes to switch from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on OLM results.

4.1.2 10 CFR Part 50 Appendix A. General Design Criterion 21 10 CFR Part 50 Appendix A. General Design Criterion 21, "Protection System Reliability and Testability," requires, in part, that plant protection systems be designed to permit periodic testing during reactor operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.

Criterion 21, Protection System Reliability and Testability. The protection system shall be designed for high functional reliability and in-service testability commensurate with the safety functions to be performed. Redundancy and independence designed into the protection system shall be sufficient to assure that (1) no single failure results in loss of the protection function and (2) removal from service of any component or channel does not result in loss of the required E-20

Enclosure to NL-24-0064 Evaluation of Proposed Changes minimum redundancy unless the acceptable reliability of operation of the protection system can be otherwise demonstrated. The protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred."

SNC proposes to use the AMS OLM methodology for FNP, HNP-1, and HNP-2 as the technical basis to support plant-specific Technical Specification changes to switch from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on OLM results. The OLM methodology is also proposed to be used to assess dynamic failure modes of pressure sensing systems. This change does not change the compliance with General Design Criterion 21 described in the FNP and HNP-2 Updated Final Safety Analysis Report (UFSAR). General Design Criterion 21 compliance is not discussed in the HNP-1 UFSARs. The licensing basis for HNP-1 was established prior to the issuance of 10 CFR Part 50 Appendix A, General Design Criteria for Nuclear Power Plants.

4.1.3 IEEE Standard 338 SNC proposes to use the AMS OLM methodology for FNP, HNP-1, and HNP-2 as the technical basis to support plant-specific Technical Specification changes to switch to time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on the OLM results for a given transmitter.

Compliance with IEEE Standard 338-1971, "IEEE Trial-Use Criteria for the Periodic Testing of Nuclear Power Generating Station Protection Systems," is described in the FNP and HNP-2 Updated Final Safety Analysis Report (UFSAR). IEEE Standard 338 compliance is not discussed in the HNP-1 UFSAR.

IEEE Standard 338-1971 contains the following requirements related to sensor testing:

5. Testing

5. 1 General Considerations. The periodic testing program shall be designed such that:

(1) The operancy (sic) of each channel and each subsystem of the protection system can be independently verified during reactor operation, except as noted in Section 5.2.

(2) Credible failures which cause functional interaction between channels can be detected.

(3) The test circuits and equipment shall not negate the protective function. The effects of any test-induced nonredundancy (sic) must be acceptable in terms of the reliability goals.

5.2 Test Intervals. An initial test interval shall be developed on the basis of the items outlined in Section 4.3. Information derived from operational data and test results (especially failure rates, MTTR, and test duration) shall be used to verify or correct the initial interval selections. During the life of the equipment, the test interval may be increased or decreased consistent with maintaining the reliability goals of the subsystem (See Appendix A).

Where it is not possible to test certain devices in the system during operation, the interval between major scheduled reactor shutdowns may be used as the test E-21

Enclosure to NL-24-0064 Evaluation of Proposed Changes interval. The designed reliability of such devices should be consistant (sic) with the test interval and the subsystem reliability goal.

Compliance with IEEE Standard 338-1971 for FNP and HNP-2 is augmented with the use of AMS-TR-0720R2-A for the condition-based calibration of transmitters in the OLM program. The AMS OLM methodology is consistent with IEEE Standard 338-2012 "IEEE Standard for Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems,"

which contains the following requirements related to sensor calibration:

5.3.3.2 On-line monitoring. On-line monitoring (OLM) techniques enable the determination of portions of an instrument channel's status during plant operation. This methodology is an acceptable input for establishing calibration frequency of those monitored portions of instrument channels without adversely affecting reliability.

Continuous monitoring shall be employed, e.g., through the plant computer.

Periodic manual testing is either a maintenance or surveillance task and is not on-line monitoring.

On-line monitoring shall ensure that setpoint calculation assumptions and the safety analysis assumptions remain valid.

4.2 Precedent The SNC license amendment request is based the NRC-approved Analysis and Measurement Services Corporation Topical Report AMS-TR-0720R2, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters" (References 1 and 2). One precedent was identified. NRC approved a license amendment request submitted by Southern Nuclear Operating Company for Vogtle Electric Generating Plant Units 1 and 2 to extend calibration intervals of nuclear plant pressure transmitters using AMS-TR-0720R2 (References 15 and 16).

4.3 No Significant Hazards Consideration Determination Analysis SNC has evaluated the proposed changes to the FNP, HNP-1, and HNP-2 Technical Specifications (TS) using the criteria in 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

The proposed changes revise the following TSs:

FNP, HNP-1, and HNP-2 TS definition 1.1 "Use and Application Definitions" The proposed changes add new Online Monitoring Program TSs, as shown below:

FNP TS 5.5.21 "Online Monitoring Program" HNP-1 TS 5.5.17 "Online Monitoring Program" HNP-2 TS 5.5.17 "Online Monitoring Program" SNC proposes to use online monitoring (OLM) methodology as the technical basis to switch from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on OLM results. Switching from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency will not create any physical changes to the plant. The use of the NRC-approved OLM methodology ensures that plant safety is maintained by demonstrating that transmitters are functioning correctly.

As required by 10 CFR 50.91 (a), the SNC analysis of the issue of no significant hazards consideration is presented below:

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Enclosure to NL-24-0064 Evaluation of Proposed Changes

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

Response: No The proposed change uses online monitoring (OLM) methodology as the technical basis to switch from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on OLM results. Switching from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency will not create any physical changes to the plant. The use of the NRC-approved OLM methodology ensures that plant safety is maintained by demonstrating that transmitters are functioning correctly.

The proposed changes do not adversely affect accident initiators or precursors, and do not alter the design assumptions, conditions, or configuration of the plant or the way the plant is operated or maintained.

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

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

Response: No The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation.

Existing calibration methods will be used when the need for transmitter calibration is determined. The change does not alter assumptions made in the safety analysis but ensures that the transmitters operate as assumed in the accident analysis. The proposed change is consistent with the safety analysis assumptions. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation.

The change does not alter assumptions made in the safety analysis but ensures that the transmitters operate as assumed in the accident analysis. The proposed change is consistent with the safety analysis assumptions. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

4.4 Conclusions In conclusion, based on the considerations discussed above, SNC concludes: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.

5 ENVIRONMENTAL CONSIDERATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20, and would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a E-23

Enclosure to NL-24-0064 Evaluation of Proposed Changes significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change. 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.

6 REFERENCES

1. Analysis and Measurement Services Corporation letter to NRC dated August 20, 2021, "Submittal of -A Version of Analysis and Measurement Services Corporation Topical Report AMS-TR-0720R2, 'Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters' (Docket No. 99902075)," (ADAMS Accession No. ML21235A493)

2. NRC Form 896, AMS Topical Report -A Verification, dated September 22, 2021 (ADAMS Accession No. ML21237A490)

3. AMS Report FNP2301 RO, "OLM Amenable Transmitters Report for Farley Units 1 & 2"

4. AMS Report HAT2301R0, "OLM Amenable Transmitters Report for Hatch Units 1 & 2"

5. AMS Report FNP2302R0, "OLM Analysis Methods and Limits Report for Farley Units 1 & 2"

6. AMS Report HAT2302R0, "OLM Analysis Methods and Limits Report for Hatch Units 1 & 2"

7. AMS Report FNP2303R0, "OLM Drift Monitoring Program Report for Farley Units 1 & 2"

8. AMS Report HAT2303R0, "OLM Drift Monitoring Program Report for Hatch Units 1 & 2"

9. AMS Procedure OLM2201, "Procedure for Online Monitoring Data Retrieval," November 2022

10. AMS Procedure OLM2202, "Procedure for Performing Online Monitoring Data Qualification and Analysis," November 2022

11. AMS Report FNP2304R0, "OLM Noise Analysis Program Report for Farley Units 1 & 2"

12. AMS Report HAT2304R0, "OLM Noise Analysis Program Report for Hatch Units 1 & 2"

13. AMS Procedure NPS1501, "Procedure for Noise Data Collection from Plant Sensors,"

March 2015

14. AMS Procedure NAR2201, "Procedure for Performing Dynamic Failure Mode Assessment Using Noise Analysis," November 2022

15. Southern Nuclear Operating Company letter NL-22-0764 to NRC dated December 21, 2022, "License Amendment Request to Revise Technical Specification 1.1 and Add 5.5.23 to Use Online Monitoring Methodology," (ADAMS Accession No. ML22355A588)

16. NRC letter to Southern Nuclear Operating Company dated June 15, 2023, ""Vogtle Electric Generating Plant, Units 1 And 2 - Issuance of Amendments Regarding Revision to Technical Specifications to Use Online Monitoring Methodology,"" (ADAMS Accession No. ML23115A149)

E-24 Joseph M Farley Nuclear Generating Plant Units 1 and 2 Proposed Technical Specification Changes (Marked-up Pages)

1.0 USE AND APPLICATION 1.1 Definitions Definitions 1.1

---

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

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.

Term ACTIONS ACTUATION LOGIC TEST AXIAL FLUX DIFFERENCE (AFD)

CHANNEL CALIBRATION Farley Units 1 and 2 Definition 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 a two section excore neutron detector.

A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel so that it responds within the required range and accuracy to known input. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensori alarm, interlock, and trip functions.

Calibration of instr nt channels with resistance temperature detecto RTD) or thermocouple sensors may consist of an inplace alitative assessment of sensor behavior and normal c libration of the remaining adjustable devices in the channel. Whenever a sensing element is replaced, the next requir d CHANNEL CALIBRATION shall include an inplace cross libration that compares the other sensing elements with the ecently installed sensing element.

The CHANNEL CALIBRA ON may be performed by means of any series of sequential, verlapping, or total channel steps, and each step must b performed within the Frequency in the Surveillance re~i'Jool'~~~"'1i~~~J,lo'l,.--i the devices included in the ste.

1.1-1 Amendment No. 226 (Unit 1)

Amendment No. 223 (Unit 2)

1.1 Definitions E-AVERAGE DISINTEGRATION ENERGY ENGINEERED SAFETY FEATURE(ESF)RESPONSE TIME INSERVICE TESTING PROGRAM LEAKAGE Farley Units 1 and 2 Definitions 1.1 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > 15 minutes, making up at least 95% of the total noniodine activity in the coolant.

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 the methodology for verification have been previously reviewed and approved by the N

, or the components have been evaluated in accordan ith an NRC approved methodology.

The INSERVICE ESTING PROGRAM is the licensee program that f ills the requirements of 10 CFR 50.55a(f).

a.

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 to not interfere with the operation of leakage detection systems; or 1.1-3 (continued)

Amendment No. 240 (Unit 1)

Amendment No. 237 (Unit 2)

1.1 Definitions PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

QUADRANT POWER TILT RA TIO (QPTR)

RATED THERMAL POWER (RTP)

REACTOR TRIP SYSTEM(RTS)RESPONSE TIME SHUTDOWN MARGIN (SOM)

Farley Units 1 and 2 Definitions 1.1 The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates and the Low Temperature Overpressure Protection System applicability temperature, 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 2821 MWt.

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 the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordan ith an NRC approved methodology.

stantaneous amount of reactivity by which the reactor is s critical or would be subcritical from its All r: a cluster control assemblies (RCCAs) are f y inserted except for the single RCCA of highest eactivity worth, which is assumed to be fully withdrawn.

However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck rod in the SOM calculation. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SOM; and 1.1-5 (continued)

Amendment No. 230 (Unit 1)

Amendment No. 227 (Unit 2)

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.20 Risk Informed Completion Time Program (continued)

If there is a high degree of confidence, based on the evidence collected, that there is no CC failure mechanism that could affect the redundant components, the RICT calculation may use nominal CC factor probability.

If a high degree of confidence cannot be established that there is no CC failure mechanism that could affect the redundant components, the RICT shall account for the increased possibility of CC failure. Accounting for the increased possibility of CC failure shall be accomplished by one of two methods. If one of the two methods listed below is not used, the Technical Specifications Front Stop shall not be exceeded.

1.

The RICT calculation shall be adjusted to numerically account for the increased possibility of CC failure, in accordance with RG 1.177, as specified in Section A-1.3.2.1 of Appendix A of the RG. Specifically, when a component fails, the CC failure probability for the remaining components shall be increased to represent the conditional failure probability due to CC failure of these components, in order to account for the possibility the first failure was caused by a CC mechanism.

OR

2.

Prior to exceeding the front stop, RMAs not already credited in the RICT calculation shall be implemented. These RMAs shall target the success of the redundant and/or diverse SSC of the failed SSC and, if possible, reduce the frequency of initiating events which call upon the function(s) performed by the failed SSCs. Documentation of RMAs shall be available for NRC review.

h.

A RICT entry is not permitted, or a RICT entry made shall be exited, for any condition involving a TS loss of function if a PRA Functionality determination that reflects the plant configuration concludes that the LCO cannot be restored without placing the TS inoperable trains in an alignment which results in a loss of functional level PRA success criteria.

Farley Units 1 and 2 5.5-18 Amendment No. 225 (Unit 1)

Amendment No. 222 (Unit 2)

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 on line 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.

Edwin I. Hatch Nuclear Power Plant Unit 1 Proposed Technical Specification Changes (Marked-up Pages)

Definitions 1.1 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 ACTIONS AVERAGE PLANAR LINEAR HEAT GENERATION RA TE (APLHGR)

CHANNEL CALIBRATION Definition ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

The APLHGR shall be applicable to a specific planar height and is equal to the sum of the LHGRs for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle at the height.

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 the entire channel, including the required senso alarm, display, and trip functions, and shall include the CHANNEL F TIONAL TEST. Calibration of instrument channels with resistanc temperature detector (RTD) or thermocouple sensors may consi of an inplace qualitative assessment of sensor behavior and normal c ibration of the remaining adjustable devices in the channel. The C NNEL CALIBRATION may be performed by means of any s Ies of sequential, overlapping, or total channel steps, and each ste must be performed within the Frequency in the Surveillance Frequency ontrol Program for the devices included in the step.

CHANNEL CHECK II be the qualitative assessment, by observation, of channel behavior d Ing operation. This determination shall include, where possible, com rison of the channel indication and status to other indications or status erived from independent instrument channels (excluding transmitters in the Online Monitoring Program)

HATCH UNIT 1 parameter.

(continued) 1.1-1 Amendment No. 303

Definitions 1.1 1.1 Definitions (continued)

MINIMUM CRITICAL POWER RA TIO (MCPR)

MODE OPERABLE -

OPERABILITY PHYSICS TESTS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

RATED THERMAL POWER (RTP)

REACTOR PROTECTION SYSTEM (RPS)

RESPONSE TIME HATCH UNIT 1 The MCPR shall be the smallest critical power ratio (CPR) that exists in the core for each class of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related INSERT:

In lieu of measurement, response time may be verified for transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

The PTLR is the unit specific docum t that provides the reactor vessel pressure and temperature limits, includ1 heatup and cooldown rates, for the current reactor vessel fluence period.

ese pressure and temperature limits shall be determined for ea fluence period in accordance with Specification 5.6.7.

RTP shall be a total reactor core heat transfer rate to coolant of 2804 MWt.

The RPS RESPONSE TIME shall be that time interval from n the monitored parameter exceeds its RPS trip setpoint at the chann sensor until de-energization of the scram pilot valve solenoids. Th response time may be measured by means of any series of sequenti overlapping, or total steps so that the entire response time is measure (continued) 1.1-5 Amendment No. 290

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 Risk Informed Completion Time Program (continued)

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 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 used to support this license amendment, 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.

HATCH UNIT 1 5.0-19a Amendment No. 319

5.5.17 INSERT 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 on line 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.

Edwin I. Hatch Nuclear Power Plant Unit 2 Proposed Technical Specification Changes (Marked-up Pages)

Definitions 1.1 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 ACTIONS AVERAGE PLANAR LINEAR HEAT GENERATION RA TE (APLHGR)

CHANNEL CALIBRATION CHANNEL CHECK Definition ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

The APLHGR shall be applicable to a specific planar height and is equal to the sum of the LHGRs for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle at the height.

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 the entire channel, including the required senso, alarm, display, and trip functions, and shall include the CHANNEL FUN NAL TEST. Calibration of instrument channels with resistance tempe ture detector (RTD) or thermocouple sensors may consist of an inp ce qualitative assessment of sensor behavior and normal calibration of th remaining adjustable devices in the channel. The CHANNEL CA IBRATION may be performed by means of any series of sequential, ov rlapping, or total channel steps, and each step must be performed wi in the Frequency in the Surveillance Frequency Control Program for e devices included in the step.

A CHANN L CHECK shall be the qualitative assessment, by observation, of channe behavior during operation. This determination shall include, where po sible, comparison of the channel indication and status to other indicatio s or status derived from independent instrument channels measur" g the same parameter.

(excluding transmitters in the Online Monitoring Program)

HATCH UNIT 2 1.1-1

( continued)

Amendment No. 248

Definitions 1.1 1.1 Definitions (continued)

EMERGENCY CORE COOLING SYSTEM (ECCS)

RESPONSE

TIME END OF CYCLE RECIRCULATION PUMP TRIP (EOC-RPT)

SYSTEM RESPONSE TIME INSERVICE TESTING PROGRAM ISOLATION SYSTEM RESPONSE TIME LEAKAGE HATCH UNIT 2 The ECCS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS initiation setpoint at the channel sensor until the ECCS 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.

SYSTEM RESPONSE TIME shall be that time interval I generation by the associated turbine stop valve limit en the turbine control valve hydraulic control oil pressure drops low the pressure switch setpoint to complete suppression of th electric arc between the fully open contacts of the recirculation pum circuit breaker. The response time may be measured by means of any s ries of sequential, overlapping, or total steps so that the entire response time is measured.

The INSERVICE TE TING PROGRAM is the licensee program that fulfills the requiremen s of 10 CFR 50.55a(f).

The ISOLATION SYST M RESPONSE TIME shall be that time interval from when the monitore parameter exceeds its isolation initiation setpoint at the channel s nsor until the isolation valves travel to their required positions. Time shall include diesel generator starting and sequence loading delays, here applicable. The response time may be measured by means of an series of sequential, overlapping, or total steps so that the entire res onse time is measured.

LEAKAGE shall be:

a.

Identified LEAKAGE

1.

LEAKAGE int as that from pump seals or valve packi d and conducted to a sump or collecting ta

2.

LEAKAGE into osphere from sources that In lieu of measurement, response time may be verified for transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

1.1-4 Amendment No. 259

Definitions 1.1 1.1 Definitions (continued)

PHYSICS TESTS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

RATED THERMAL POWER (RTP)

REACTOR PROTECTION SYSTEM (RPS)

RESPONSE TIME SHUTDOWN MARGIN (SOM)

STAGGERED TEST BASIS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a.

Described in Chapter 14, Initial Tests and Operation, of the FSAR;

b.

Authorized under the provisions of 10 CFR 50.59; or

c.

Otherwise approved by the Nuclear Regulatory Commission.

The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.7.

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2804 MWt.

The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measur SOM shall be the amount of reactivity by which the reactor is su 1tical or would be subcritical throughout the operating cycle assu

• g that:

a.

The reactor is xenon free;

b.

The moderator temperature is~ 6 reactive state; and In lieu of measurement, response time may be verified for transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

(continued)

HATCH UNIT 2 1.1-6 Amendment No. 235

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16

<(

Risk Informed Completion Time Program (continued)

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

e.

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.

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 used to support this license amendment, 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.

~

NewTru17here HATCH UNIT 2 5.0-19a Amendment No. 264

5.5.17 INSERT 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 on line 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.

Joseph M Farley Nuclear Generating Plant Units 1 and 2 Revised Technical Specification Pages

1.0 USE AND APPLICATION 1.1 Definitions Definitions 1.1

---

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

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.

Term ACTIONS ACTUATION LOGIC TEST AXIAL FLUX DIFFERENCE (AFD)

CHANNEL CALIBRATION Farley Units 1 and 2 Definition 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 a two section excore neutron detector.

A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel so that it responds within the required range and accuracy to known input. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor (excluding transmitters in the Online Monitoring Program), alarm, interlock, and trip functions.

Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.

The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

1.1-1 Amendment No.

Amendment No.

(Unit 1)

(Unit 2)

1.1 Definitions E-AVERAGE DISINTEGRATION ENERGY ENGINEERED SAFETY FEATURE(ESF)RESPONSE TIME INSERVICE TESTING PROGRAM LEAKAGE Farley Units 1 and 2 Definitions 1.1 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > 15 minutes, making up at least 95% of the total noniodine activity in the coolant.

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 the methodology for verification have been previously reviewed and approved by the NRC (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.

The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

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 to not interfere with the operation of leakage detection systems; or 1.1-3 Amendment No.

Amendment No.

(continued)

(Unit 1)

(Unit 2)

1.1 Definitions PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

QUADRANT POWER TILT RA TIO (QPTR)

RA TED THERMAL POWER (RTP)

REACTOR TRIP SYSTEM(RTS)RESPONSE TIME SHUTDOWN MARGIN (SOM)

Farley Units 1 and 2 Definitions 1.1 The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates and the Low Temperature Overpressure Protection System applicability temperature, 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 2821 MWt.

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 the methodology for verification have been previously reviewed and approved by the NRC (including transmitters in the Online Monitoring Program), or the components have been evaluated in accordance with an NRC approved methodology.

SOM 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.

However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck rod in the SOM calculation. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SOM; and 1.1-5 Amendment No.

(Unit 1)

Amendment No.

(Unit 2)

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.20 5.5.21 Risk Informed Completion Time Program (continued)

If there is a high degree of confidence, based on the evidence collected, that there is no CC failure mechanism that could affect the redundant components, the RICT calculation may use nominal CC factor probability.

If a high degree of confidence cannot be established that there is no CC failure mechanism that could affect the redundant components, the RICT shall account for the increased possibility of CC failure. Accounting for the increased possibility of CC failure shall be accomplished by one of two methods. If one of the two methods listed below is not used, the Technical Specifications Front Stop shall not be exceeded.

1.

The RICT calculation shall be adjusted to numerically account for the increased possibility of CC failure, in accordance with RG 1.177, as specified in Section A-1.3.2.1 of Appendix A of the RG. Specifically, when a component fails, the CC failure probability for the remaining components shall be increased to represent the conditional failure probability due to CC failure of these components, in order to account for the possibility the first failure was caused by a CC mechanism.

OR

2.

Prior to exceeding the front stop, RMAs not already credited in the RICT calculation shall be implemented. These RMAs shall target the success of the redundant and/or diverse SSC of the failed SSC and, if possible, reduce the frequency of initiating events which call upon the function(s) performed by the failed SSCs. Documentation of RMAs shall be available for NRC review.

h.

A RICT entry is not permitted, or a RICT entry made shall be exited, for any condition involving a TS loss of function if a PRA Functionality determination that reflects the plant configuration concludes that the LCO cannot be restored without placing the TS inoperable trains in an alignment which results in a loss of functional level PRA success criteria.

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.

(continued)

Farley Units 1 and 2 5.5-18 Amendment No.

Amendment No.

(Unit 1)

(Unit 2)

Programs and Manuals 5.5 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 on line 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 on line 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.

Farley Units 1 and 2 5.5-19 Amendment No.

Amendment No.

(Unit 1)

(Unit 2)

Edwin I. Hatch Nuclear Power Plant Unit 1 Revised Technical Specification Pages

Definitions 1.1 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 ACTIONS AVERAGE PLANAR LINEAR HEAT GENERATION RA TE (APLHGR)

CHANNEL CALIBRATION Definition ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

The APLHGR shall be applicable to a specific planar height and is equal to the sum of the LHGRs for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle at the height.

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 the entire channel, including the required sensor (excluding transmitters in the Online Monitoring Program), alarm, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace 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, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

CHANNEL CHECK 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)

HATCH UNIT 1 1.1-1 Amendment No.

Definitions 1.1 1.1 Definitions (continued)

MINIMUM CRITICAL POWER RA TIO (MCPR)

MODE OPERABLE-OPERABILITY PHYSICS TESTS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

RATED THERMAL POWER (RTP)

REACTOR PROTECTION SYSTEM (RPS)

RESPONSE TIME HATCH UNIT 1 The MCPR shall be the smallest critical power ratio (CPR) that exists in the core for each class of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a.

Described in Section 13.6, Startup and Power Test Program, of the FSAR;

b.

Authorized under the provisions of 10 CFR 50.59; or

c.

Otherwise approved by the Nuclear Regulatory Commission.

The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, 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. 7.

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2804 MWt.

The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. 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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

1.1-5 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.17 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 AMSTR-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.

HATCH UNIT 1 5.0-19b Amendment No.

Edwin I. Hatch Nuclear Power Plant Unit 2 Revised Technical Specification Pages

Definitions 1.1 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 ACTIONS AVERAGE PLANAR LINEAR HEAT GENERATION RA TE (APLHGR)

CHANNEL CALIBRATION Definition ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

The APLHGR shall be applicable to a specific planar height and is equal to the sum of the LHGRs for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle at the height.

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 the entire channel, including the required sensor (excludimg transmitters in the Online Monitoring Program), alarm, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace 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, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

CHANNEL CHECK 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.

HATCH UNIT 2 1.1-1

( continued)

Amendment No.

Definitions 1.1 1.1 Definitions (continued)

EMERGENCY CORE COOLING SYSTEM (ECCS)

RESPONSE

TIME END OF CYCLE RECIRCULATION PUMP TRIP (EOC-RPT)

SYSTEM RESPONSE TIME INSERVICE TESTING PROGRAM ISOLATION SYSTEM RESPONSE TIME LEAKAGE HATCH UNIT 2 The ECCS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS initiation setpoint at the channel sensor until the ECCS 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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

The EOC-RPT SYSTEM RESPONSE TIME shall be that time interval from initial signal generation by the associated turbine stop valve limit switch or from when the turbine control valve hydraulic control oil pressure drops below the pressure switch setpoint to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker. 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.

The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).

The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves travel to their required positions. 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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

LEAKAGE shall be:

a.

Identified LEAKAGE

b.

1.

LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or

2.

LEAKAGE into the drywell 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; Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE; (continued) 1.1-4 Amendment No.

Definitions 1.1 1.1 Definitions (continued)

PHYSICS TESTS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

RATED THERMAL POWER (RTP)

REACTOR PROTECTION SYSTEM (RPS)

RESPONSE TIME SHUTDOWN MARGIN (SOM)

STAGGERED TEST BASIS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a.

Described in Chapter 14, Initial Tests and Operation, of the FSAR;

b.

Authorized under the provisions of 10 CFR 50.59; or

c.

Otherwise approved by the Nuclear Regulatory Commission.

The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, 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. 7.

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2804 MWt.

The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. 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 transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC SOM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:

a.

The reactor is xenon free;

b.

The moderator temperature is~ 68°F, corresponding to the most reactive state; and

c.

All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.

With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SOM.

A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

HATCH UNIT 2 1.1-6

( continued)

Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.17 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.

HATCH UNIT 2 5.0-19b Amendment No.

Joseph M Farley Nuclear Generating Plant Units 1 and 2 Technical Specification Bases Mark-up Pages (Information Only)

BASES SURVEILLANCE REQUIREMENTS SR 3.3.1.9 (continued)

RTS Instrumentation B 3.3.1 The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.1.10 CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the unit specific setpoint methodology. The "as found" and "as left" data have been evaluated to ensure consistency with (i.e.,

bounded by) the drift allowance used in the setpoint methodology.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by two N es. Note 1 states that neutron detectors are excluded from the CHAN EL CALIBRATION where applicable.

The CHANNEL CALIBRATIO for the power range neutron detectors consists of a normalization of he detector outputs based on an incore/excore cross-calibrati (SR 3.3.1.9). In addition, the CHANNEL CALIBRATION for the power range neutron detector outputs includes normalization of the channel utput based on a power calorimetric (SR 3.3.1.2) performed above 15 o RTP. The CHANNEL CALIBRATION for the intermediate range neutr n detector outputs includes normalization of the high flux bistable base on a power calorimetric. The CHANNEL CALIBRATION for the sourc range neutron detectors consists of obtaining new detector plate u and preamp discriminator curves after a detector is replaced. This S rveillance is not required for the NIS power range detectors for entry int MODE 2 or 1, and is not required for the NIS intermediate range det tors for entry into MODE 2, because the unit must be in at least MO E 2 to perform the test for the intermediate range detectors and MODE 1 for the power range detectors. Note 2 Alternately, the Frequency for checking the calibration of pressure, level, and flow transmitters may be determined in accordance with the Online Monitoring Program implemented in accordance with AMS-TR-0720R2-A (Ref. 24) and TS 5.5.21, Online Monitoring Program.

(continued)

Farley Units 1 and 2 B 3.3.1-56 Revision 52

BASES SURVEILLANCE REQUIREMENTS SR 3.3.1.14 (continued) types must be demonstrated by test.

RTS Instrumentation B 3.3.1 WCAP-14036-P-A, Revision 1, "Eli nation of Periodic Protection Channel Response Time Tests," (R f. 19) provides the basis and methodology for using allocated si al processing and actuation logic response times in the overall verifi tion of the protection system channel response time. The alloc tions for the sensor, signal conditioning and actuation logic r ponse times must be verified prior to placing the component in operati al service and re-verified following maintenance that may adversely ffect response time. In general, electric repair work does not imp ct response time provided the parts used for repair are of the same pe and value. Specific components identified in the WCAP may be placed without verification testing.

One example where time resp se could be affected is replacing the sensing assembly of a transmi er.

The response time may be ve ified for components that replace the components that were previo sly evaluated in Ref. 18 and Ref. 19, provided that the component have been evaluated in accordance with the NRC approved methodo gy as discussed in Attachment 1 to TSTF-569, "Methodology to Elimi te Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing," (Ref.

24).

The Surveillance Frequen y is controlled under the Surveillance Frequency Control Progr SR 3.3.1.14 is modified a Note stating that neutron detectors are excluded from RTS RE ONSE TIME testing. This Note is necessary because of the difficulty In generating an appropriate detector input signal. Excluding the d teeters is acceptable because the principles of detector operation ens re a virtually instantaneous response.

Alternately, the use of the allocated RTS RESPONSE TIME for transmitters in the Online Monitoring Program is supported by the performance of ONLINE MONITORING using the 'noise analysis' technique to detect dynamic failures modes that can affect transmitter response time.

Farley Units 1 and 2

3.

FSAR, Chapter 15.

4.

Joseph M. Farley Nuclear Power Plant Unit 1 (2) Precautions, Limitations and Setpoints U-266647 (U-280912).

( continued)

B 3.3.1-59 Revision 99

BASES REFERENCES (continued)

Insert:

RTS Instrumentation B 3.3.1

17. Westinghouse Technical Bulletin, NSD-TB-92-03-R1, "Undervoltage Trip Protection."

18. WCAP-13632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," Jan., 1996.

19. WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," Oct., 1998.

20. WCAP 12925, Median Signal Selector (MSS).

21. WCAP 13807/13808, Elimination of Feedwater Flow trip via Implementation of MSS.

22. SNC Calculation E-35.1A & E-35.2A.

23. Regulatory Guide 1.105, Revision 3, "Setpoints for Safety-Related Instrumentation."

24. Attachment 1 to TSTF-569, "Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only)

Response Time Testing."

24. AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters."

Farley Units 1 and 2 B 3.3.1-61 Revision 99

BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.2.7 ESFAS Instrumentation B 3.3.2 SR 3.3.2.7 is the performance of a CHANNEL CALIBRATION.

CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the unit specific setpoint methodology. The "as found" and "as left" data have been evaluated to ensure consistency with (i.e., bounded by) the drift allowance used in the setpoint methodology.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program This SR is modified by a No stating that this test should include verification that the time co tants are adjusted to the prescribed values where applicable.

SR 3.3.2.8 SR 3.3.2.8 is the perform ce of a SLAVE RELAY TEST. The SLAVE RELAY TEST is t e energizing of the slave relays. Contact operation is verified in on of two ways. Actuation equipment that may be operated in the d sign mitigation MODE is either allowed to function, or is placed in condition where the relay contact operation can be verified without o eration of the equipment. Actuation equipment that may not e operated in the design mitigation MODE is prevented from operati by the SLAVE RELAY TEST circuit or is Alternately, the Frequency for checking the calibration of pressure, level, and flow transmitters may be determined in accordance with the Online Monitoring Program implemented in accordance with AMS-TR-0720R2-A (Ref. 21) and TS 5.5.21, Online Monitoring Program.

Farley Units 1 and 2 relay testing is normally conducted during refueling to minimize the potential for plant transients and unnecessary challenges to plant equipment.

(continued)

B 3.3.2-48 Revision 94

BASES SURVEILLANCE REQUIREMENTS

( continued)

Insert:

SR 3.3.2.9 ESFAS Instrumentation B 3.3.2 This SR ensures the individual channel ESF RESPONSE TIMES are less than or equal to the maximum values assumed in the accident analysis. Response Time testing acceptance criteria are included in the FSAR, Table 7.3-16 (Ref. 13). Individual component response times are not typically modeled in the analyses. The analyses model the overall or total elapsed time, from the point at which the parameter exceeds the Trip Setpoint value at the sensor, to the point at which the equipment reaches the required functional state (e.g., pumps at rated discharge pressure, valves in full open or closed position).

For channels that include dynamic transfer functions (e.g., lag, lead/lag, rate/lag, etc.), the response time test may be performed with the transfer functions set to one or with the time constants set to their nominal value. The test results must be compared to properly defined acceptance criteria.

Response time may be verified by actual response time tests in any Alternately, the use of the allocated ESF RESPONSE TIME for transmitters in the Online Monitoring Program is supported by the performance of ONLINE MONITORING using the 'noise analysis' technique to detect dynamic failures modes that can affect transmitter response time.

Farley Units 1 and 2 tests), (2) in place, ons

, or offsite (e.g., vendor) test measurements, or (3) utilizing vendor en

• eering specifications.

WCAP-13632-P-A, Revision, "Elimination of Pressure Sensor Response Time Testing Requ1 ments," (Ref. 14) provides the basis and methodology for using alloc ed sensor response times in the overall verification of the channel s onse time for specific sensors identified in the WCAP. Respons e verification for other sensor types must be demonstrated by test.

WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," (Ref. 15) provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time. The allocations for the sensor, signal processing and actuation logic response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time. In general, (continued)

B 3.3.2-49 Revision 94

BASES REFERENCES

( continued)

Insert:

3.

FSAR, Chapter 15.

ESFAS Instrumentation B 3.3.2

4.

Joseph M. Farley Nuclear Power Plant Unit 1 (2) Precautions, Limitations, and Setpoints U-26664 7 (U-280912).

5.

IEEE-279-1971.

6.

WCAP 13751, Rev. 1, Westinghouse Setpoint Methodology for Protection Systems Farley Nuclear Plant Units 1 and 2.

7.

10 CFR 50.49.

8.

WCAP 13751 Rev. 0, Westinghouse Setpoint Methodology for Protection Systems SNOC Farley Nuclear Plant Units 1 and 2.

9.

Not used.

10.

WCAP-10271-P-A, Supplement 2, Rev. 1, "Updated Approved Version," June 1990.

11.

WCAP-14333-P-A, Revision 1, "Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times,"

21. AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters."

of Pressure Sensor

., 1996.

Farley Units 1 and 2

15.

16.

CAP-14036-P-A, Revision 1, "Elimination of Periodic Protection hannel Response Time Tests," Oct. 1998.

UREG-1218, April 1988.

17.

~-181007 Reactor Protection System FSD.

18.

Westinghouse Functional Diagrams U-166231 thru U-166245.

19.

WCAP-16294-NP-A, Rev. 1, "Risk-Informed Evaluation of Changes to Technical Specification Required Action Endstates for Westinghouse NSSS PWRs," June 2010. to TSTF-569, "Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only)

Response Time Testing."

B 3.3.2-51 Revision 99

Insert:

BASES SURVEILLANCE REQUIREMENTS REFERENCES SR 3.3.3.2 (continued)

PAM Instrumentation B 3.3.3 The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

1.

A-181866 Unit 1 RG 1.97 Co A-204866 Unit 2 RG 1.97 Com iance Review NRC SER for FNP RG 1.97 Com to McDonald, 2/12/87.

2.

Regulatory Guide 1.97.

3.

NUREG-0737, Supplement 1, "TMI Actio Insert:

Alternately, the Frequency for checking the calibration of pressure, level, and flow transmitters may be determined in accordance with the Online Monitoring Program implemented in accordance with AMS-TR-0720R2-A (Ref. 4) and TS 5.5.21, Online Monitoring Program.

4. AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters."

Farley Units 1 and 2 B 3.3.3-16 Revision 52

Insert:

ote Shutdown System B 3.3.4

2. AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters."

REQUIREMENTS

( continued)

REFERENCES Insert:

EL CALIBRATION is a complete check of the monitoring instru ent loop and the sensor. The test verifies that the channel respo ds to a measured parameter within the necessary range and cy.

urveillance Frequency is controlled under the Surveillance Fre uency Control Program.

Alternately, the Frequency for checking the calibration of pressure, level, and flow transmitters may be determined in accordance with the Online Monitoring Program implemented in accordance with AMS-TR-0720R2-A (Ref. 2) and TS 5.5.21, Online Monitoring Program.

Farley Units 1 and 2 B 3.3.4-6 Revision 52 Edwin I. Hatch Nuclear Power Plant Unit 1 Technical Specification Bases Mark-up Pages (Information Only)

BASES APPLICABLE SAFETY ANALYSES LCO, and APPLICABILITY

( continued)

RPS Instrumentation B 3.3.1.1 the specified Allowable Value, where appropriate. The setpoint is calibrated consistent with applicable setpoint methodology assumptions (nominal trip setpoint). Each channel must also respond within its assumed response time, where appropriate.

Allowable Values are specified for each RPS Function specified in the Table. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the actual setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setpoint is not within its r ed Allowable Value.

Trip setpoints are those pred ermined values of output at which an action should take place. Th setpoints are compared to the actual process parameter (e.g., rea tor vessel water level), and when the measured output value of th process parameter exceeds the setpoint, the associated de ce (e.g., trip unit) changes state. The analytic limits are derived f m the limiting values of the process parameters obtained from e safety analysis. The Allowable Values are derived from the anal ic limits, corrected for calibration, process, and some of the instrum terrors.

The trip setpoints are th n determined accounting for the remaining instrument errors (e.g., rift). The trip setpoints derived in this manner provide adequate prote tion because instrumentation uncertainties, process effects, calibr ion tolerances, instrument drift, and severe environmental effects for channels that must function in harsh Alternately, the Frequency for checking the calibration of pressure, level, and flow transmitters may be determined in accordance with the Online Monitoring Program implemented in accordance with AMS-TR-0720R2-A (Ref. 21) and TS 5.5.17, Online Monitoring Program.

HATCH UNIT 1 fi I

i u

I ns ar r u e o e In MODES or other specified conditions specified in the Table, which may require an RPS trip to mitigate the consequences of a design basis accident or transient. To ensure a reliable scram function, a combination of Functions are required in each MODE to provide primary and diverse initiation signals. The only MODES specified in Table 3.3.1.1-1 are MODES 1 (which encompasses~ 27.6% RTP) and 2, and MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies. No RPS Function is required in MODES 3 and 4 since all control rods are fully inserted and the Reactor Mode Switch Shutdown Position control rod withdrawal block (continued)

B 3.3-3 REVISION 36

BASES SURVEILLANCE REQUIREMENTS SR 3.3.1.1.13 RPS Instrumentation B 3.3.1.1 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the plant specific setpoint methodology. For MSIV - Closure, SDV Water Level - High (Float Switch), and T Closure Functions, this SR also includes a physical inspection and ctuation of the switches. For the APRM Simulated Thermal Power High Function, this SR also includes calibrating the associate recirculation loop flow channel.

Note 1 states that neutron detec rs are excluded from CHANNEL CALIBRATION because they ar passive devices, with minimal drift, and because of the difficulty of 1mulating a meaningful signal.

Changes in neutron detector s sitivity are compensated for by performing the calorimetric car ration (SR 3.3.1.1.2) and the LPRM calibration against the TIPs (S 3.3.1.1.8). A second Note is provided that requires the IRM SRs to e performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of entering MODE 2 from MODE 1. Tes ng of the MODE 2 IRM Functions cannot be performed in MO E 1 without utilizing jumpers, lifted leads or movable links. This Note allows entry into MODE 2 from MODE 1 if the associated Frequency i not met per SR 3.0.2.

Twelve hours is based on perating experience and in consideration of providing a reasonable ime in which to complete the SR.

The Surveillance Freque cy is controlled under the Surveillance Frequency Control Prog am.

SR 3.3.1.1.14 (Not used.)

Alternately, the Frequency for checking the calibration of pressure, level, and flow transmitters may be determined in accordance with the Online Monitoring Program implemented in accordance with AMS-TR-0720R2-A (Ref. 21) and TS 5.5.17, Online Monitoring Program.

(continued)

HATCH UNIT 1 B 3.3-28 REVISION 85

BASES SURVEILLANCE REQUIREMENTS

( continued)

Insert:

SR 3.3.1.1.15 RPS Instrumentation B 3.3.1.1 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The functional testing of control rods (LCO 3.1.3), and SDV vent and drain valves (LCO 3.1.8), overlaps this Surveillance to provide complete testing of the assumed safety function.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

The LOGIC SYSTEM FUNCTIONAL TEST for APRM Function 2.e simulates APRM and OPRM trip conditions at the two-out-of-four voter channel inputs to check all combinations of two tripped inputs to the two-out-of-four logic in the voter channels and APRM related redundant RPS relays.

SR 3.3.1.1.16 This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis. This test may be performed in one measurement or in overlapping segments, with verification that all components are tested. The RPS RESPONSE TIME acceptance criteria are included in Reference 10.

RPS RESPONSE IME for APRM Two-out-of-Four Voter Function 2.e includes the output relays of the voter and the associated RPS relays and contactors. (T e digital portions of the APRM and two-out-of-four voter channels are xcluded from RPS RESPONSE TIME testing because self-testing and calibration check the time base of the digital electronics.) Confir ation of the time base is adequate to assure required response ti es are met. Neutron detectors are excluded from RPS RESPON E TIME testing because the principles of detector operation vir ually ensure an instantaneous response time.

The Surveillance Fre ency is controlled under the Surveillance Frequency Control Pr ram.

Alternately, the use of the allocated RPS RESPONSE TIME for transmitters in the Online Monitoring Program is supported using the 'noise analysis' technique implemented in accordance with AMS-TR-0720R2-A (Ref. 21) to detect dynamic failures modes that can affect transmitter response time.

(continued)

HATCH UNIT 1 B 3.3-29 REVISION 85

BASES REFERENCES

( continued)

Insert:

14.

RPS Instrumentation B 3.3.1.1 NEDO-31960-A, Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology,"

November 1995.

15.

NEDE-33766P-A, "GEH Simplified Stability Solution (GS3),"

March 2015.

16.

NEDO-3241 OP-A, Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Ill Stability Trip Function,"

November 1997.

17.

Letter, L.A. England (BWROG) to M.J. Virgilio, "BWR Owners' Group Guidelines for Stability Interim Corrective Action,"

June 6, 1994.

18.

Not used.

19.

GE Letter NSA 02-250, "Plant Hatch IRM Technical Specifications," April 19, 2002.

20.

004N6160, "Edwin I. Hatch Nuclear Plant, Units 1 and 2, TRACG-LOCA Loss-of-Coolant Accident Analysis," Revision 1, October 2019.

21. AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters."

HATCH UNIT 1 B 3.3-32 REVISION 110

BASES PAM Instrumentation B 3.3.3.1 SURVEILLANCE REQUIREMENTS

( continued) 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. The Note is based upon a NRC Safety Evaluation Report (Ref. 2) which concluded that the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> testing allowance does not significantly reduce the probability of properly monitoring post accident parameters, when necessary.

HATCH UNIT 1 Insert:

SR 3.3.3.1.1 Performance of the CHANNEL CHECK once every 31 days ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel against a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the plant staff, based on a Alternately, the Frequency for checking the calibration of pressure, level, and flow transmitters may be determined in accordance with the Online Monitoring Program implemented in accordance with AMS-TR-0720R2-A (Ref. 4) and TS 5.5.17, Online Monitoring Program.

SR 3.3.3.1.2 CHANNEL CALIBRATION is a c plete check of the instrument loop, including the sensor. The test v rifies the channel responds to measured parameter with the n cessary range and accuracy.

The Surveillance Frequency

• Frequency Control Program.

B 3.3-68 ntrolled under the Surveillance (continued)

REVISION 69

PAM Instrumentation B 3.3.3.1 BASES (continued)

REFERENCES HATCH UNIT 1 Insert:

1.

Regulatory Guide 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Revision 2, December 1980.

2.

NRC Safety Evaluation Report, "Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, Conformance to Regulatory Guide 1.97,"

dated July 30, 1985.

3.

NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

4. AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters."

B 3.3-69 REVISION 69

BASES Remote Shutdown System B 3.3.3.2 SURVEILLANCE REQUIREMENTS SR 3.3.3.2.1 (continued)

Insert:

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

4. AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters."

nction. This nd locally, hutdown

._._._._,_,_,_,_"""""'~.,,.,_,._..,~,._,:,~..,,.......,,....c:1~"""'1'1r.,,,,r,~~~~~~ by performance ofi a continuity check, or, in the case of the DG controls, the routine Su eillances of LCO 3.8.1 (since local control is utilized during the pe ormance of some of the Surveillances of LCO 3.8.1 ).

This will ens e that if the control room becomes inaccessible, the plant can be laced and maintained in MODE 3 from the remote shutdown p el and the local control stations. The Surveillance Frequency i controlled under the Surveillance Frequency Control Program.

CHANN L CALIBRATION is a complete check of the instrument loop and the ensor. The test verifies the channel responds to measured param er values with the necessary range and accuracy.

The S rveillance Frequency is controlled under the Surveillance Freq ncy Control Program.

REFERENCES

1.

Technical Requireme ts Manual, Table T6.0-1.

NRC No.93-102, "Fin I Policy Statement on Technical Specification Improve ents," July 23, 1993.

Alternately, the Frequency for checking the calibration of pressure, level, and flow transmitters may be determined in accordance with the Online Monitoring Program implemented in accordance with AMS-TR-0720R2-A (Ref. 4) and TS 5.5.17, Online Monitoring Program.

HATCH UNIT 1 B 3.3-74 REVISION 69

BASES ATWS-RPT Instrumentation B 3.3.4.2 SURVEILLANCE REQUIREMENTS SR 3.3.4.2.3 (continued)

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.4.2.4 The LOGIC SYSTEM FUN NAL TEST demonstrates the Alternately, the Frequency for checking the calibration of pressure, level, and flow transmitters may be determined in accordance with the Online Monitoring Program implemented in accordance with AMS-TR-0720R2-A (Ref. 4) and TS 5.5.17, Online Monitoring Program.

REFERENCES Insert:

1.

FSAR, Section 7.23.

2.

GENE-770-06-1, "Bases for Changes To Surveillance Test Intervals and Allowed Out-of-Service Times For Selected Instrumentation Technical Specifications," February 1991.

3.

NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

4. AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters."

HATCH UNIT 1 B 3.3-91 REVISION 69

BASES ECCS Instrumentation B 3.3.5.1 SURVEILLANCE REQUIREMENTS SR 3.3.5.1.2 (continued)

REFERENCES HATCH UNIT 1 The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.5.1.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the plant specific setpoint methodology.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.5.1.5 The LOGIC SYSTEM FUNCTIO L TEST demonstrates the OPERABILITY of the required init1 tion logic for a specific channel.

Alternately, the Frequency for checking the calibration of pressure, level, and flow transmitters may be determined in accordance with the Online Monitoring Program implemented in accordance with AMS-TR-0720R2-A (Ref. 5) and TS 5.5.17, Online Monitoring Program.

1.

FSAR, Section 4.8.

2.

FSAR, Section 6.5.

3.

Unit 2 FSAR, Chapter 15.

(Not used)

Insert:

5. AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters."

B3.3-123 REVISION 110

BASES SURVEILLANCE REQUIREMENTS

( continued)

Primary Containment Isolation Instrumentation B 3.3.6.1 SR 3.3.6.1.3, SR 3.3.6.1.4, and SR 3.3.6.1.5 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the plant specific setpoint methodology.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Alternately, the Frequency for checking the calibration of pressure, level, and flow transmitters may be determined in accordance with the Online Monitoring Program implemented in accordance with AMS-TR-0720R2-A (Ref. 7) and TS 5.5.17, Online Monitoring Program.

REFERENCES HATCH UNIT 1

1.

2.

3.

4.

5.

6.

Insert:

FSAR, Section 5.2.

Unit 2 FSAR, Chapter 15.

FSAR, Section 3.8.3.

NEDC-31677P-A, "Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation,"

July 1990.

NEDC-30851 P-A Supplement 2, "Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation," March 1989.

NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

7. AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters."

Edwin I. Hatch Nuclear Power Plant Unit 2 Technical Specification Bases Mark-up Pages (Information Only)

BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY

( continued)

RPS Instrumentation B 3.3.1.1 the specified Allowable Value, where appropriate. The setpoint is calibrated consistent with applicable setpoint methodology assumptions (nominal trip setpoint). Each channel must also respond within its assumed response time, where appropriate.

Allowable Values are specified for each RPS Function specified in the Table. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the actual setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable.

channel is inoperable if its actual trip setpoint is not within its r I ed Allowable Value.

Trip setpoints are those prede ermined values of output at which an action should take place. Th setpoints are compared to the actual process parameter (e.g., rea tor vessel water level), and when the measured output value of th process parameter exceeds the setpoint, the associated de ce (e.g., trip unit) changes state. The analytic limits are derived f m the limiting values of the process Alternately, the Frequency for checking the calibration of pressure, level, and flow transmitters may be determined in accordance with the Online Monitoring Program implemented in accordance with AMS-TR-0720R2-A (Ref. 24) and TS 5.5.17, Online Monitoring Program.

HATCH UNIT 2 process effects, calibration tolerances, instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50 49) are accounted for.

The OPERABILITY of scram pilot valves and associated solenoids, backup scram valves, and SDV valves, described in the Background section, are not addressed by this LCO.

The individual Functions are required to be OPERABLE in the MODES or other specified conditions specified in the Table, which may require an RPS trip to mitigate the consequences of a design basis accident or transient. To ensure a reliable scram function, a combination of Functions are required in each MODE to provide primary and diverse initiation signals. The only MODES specified in Table 3.3.1.1-1 are MODES 1 (which encompasses~ 27.6% RTP) and 2, and MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies. No RPS Function is required in MODES 3 and 4 since all control rods are fully inserted and the Reactor Mode Switch Shutdown Position control rod withdrawal block

( continued)

B 3.3-3 REVISION 42

BASES SURVEILLANCE REQUIREMENTS SR 3.3.1.1.11 (continued)

RPS Instrumentation B 3.3.1.1 POWER is ~ 27.6% RTP. This involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint. Because main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from turbine first stage pressure), the main turbine bypass valves must remain closed during the calibration at THERMAL POWER ~ 27.6% RTP to ensure that the calibration is Alternately, the Frequency for checking the calibration of pressure, level, and flow transmitters may be determined in accordance with the Online Monitoring Program implemented in accordance with AMS-TR-0720R2-A (Ref. 24) and TS 5.5.17, Online Monitoring Program.

HATCH UNIT 2 the bypass channel can e placed in the conservative condition (nonbypass). If placed i the nonbypass condition (Turbine Stop Valve - Closure and Turb e Control Valve Fast Closure, Trip Oil Pressure - Low Functions re enabled), this SR is met and the channel is considered OP ABLE.

The Surveillance Frequenc *s controlled under the Surveillance Frequency Control Program.

SR 3.3.1.1.13 A CHANNEL CALIBRATION is complete check of the instrument loop and the sensor. This test ve ifies that the channel responds to the measured parameter within th necessary range and accuracy.

CHANNEL CALIBRATION leaves e channel adjusted to account for instrument drifts between succes

• calibrations, consistent with the plant specific setpoint methodology. For MSIV - Closure, SDV Water Level - High (Float Switch), and TSV - Closure Functions, this SR also includes a physical inspection and actuation of the switches. For the APRM Simulated Thermal Power - High Function, this SR also includes calibrating the associated recirculation loop flow channel.

Note 1 states that neutron detectors are excluded from CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal.

Changes in neutron detector sensitivity are compensated for by performing the calorimetric calibration (SR 3.3.1.1.2) and the LPRM calibration against the TIPs (SR 3.3.1.1.8). A second Note is provided that requires the IRM SRs (continued)

B 3.3-28 REVISION 79

BASES SURVEILLANCE REQUIREMENTS Insert:

SR 3.3.1.1.16 (continued)

RPS Instrumentation B 3.3.1.1 analysis. This test may be performed in one measurement or in overlapping segments, with verification that all components are tested. The RPS RESPONSE TIME acceptance criteria are included in Reference 10.

RPS RESPONSE TIME for APRM two-out-of-four Voter Function 2.e includes the outp relays of the voter and the associated RPS relays and contactors. (

e digital portions of the APRM and two-out-of-four voter channels are excluded from RPS RESPONSE TIME testing because self-testin and calibration check the time base of the digital electronics.) Confi ation of the time base is adequate to assure required response t mes are met. Neutron detectors are excluded from RPS RESPO E TIME testing because the principles of detector operation v rtually ensure an instantaneous response time.

The Note allows neuron detectors to be excluded from RPS RESPONSE TIME t ting because the principles of detector operation virtually ensure an in tantaneous response time.

The Surveillance Fre uency is controlled under the Surveillance Frequency Control Pr gram.

Alternately, the use of the allocated RPS RESPONSE TIME for transmitters in the Online Monitoring Program is supported using the 'noise analysis' technique implemented in accordance with AMS-TR-0720R2-A (Ref. 24) to detect dynamic failures modes that can affect transmitter response time.

HATCH UNIT 2 I

S POWER, as indicated by APRM Simulated Thermal Power, is

~ 25% RTP and core flow, as indicated by recirculation drive flow, is

< 60% rated core flow. This normally involves confirming the bypass setpoints. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint. The actual Surveillance ensures that the OPRM Upscale Function is enabled (not bypassed) for the correct values of APRM Simulated Thermal Power and recirculation drive flow. Other Surveillances ensure that the APRM Simulated Thermal Power and recirculation flow properly correlate with THERMAL POWER and core flow, respectively.

If any bypass setpoint is nonconservative (i.e., the OPRM Upscale Function is bypassed when APRM Simulated Thermal Power is

~ 25% and recirculation drive flow is < 60% rated), then the affected channel is considered inoperable for the OPRM Upscale Function.

(continued)

B 3.3-30 REVISION 79

BASES REFERENCES

( continued)

HATCH UNIT 2 Insert:

14.

15.

16.

17.

18.

19.

20.

21.

22.

23.

RPS Instrumentation B 3.3.1.1 NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

NEDO-31960-A, Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology,"

November 1995.

NEDE-33766P-A, "GEH Simplified Stability Solution (GS3),"

March 2015.

NEDO-3241 OP-A, Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Ill Stability Trip Function,"

November 1997.

Letter, L.A. England (BWROG) to M.J. Virgilio, "BWR Owners' Group Guidelines for Stability Interim Corrective Action,"

June 6, 1994.

NEDO-32291-A, Supplement 1, "System Analyses for the Elimination of Selected Response Time Testing Requirements," October 1999.

Not used.

GE Letter NSA 02-250, "Plant Hatch IRM Technical Specifications," April 19, 2002.

Not used.

004N6160, "Edwin I. Hatch Nuclear Plant Units 1 and 2, TRACG-LOCA Loss-of-Coolant Accident Analysis," Revision 1, October 2019.

24. AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters."

B 3.3-32 REVISION 122

BASES SURVEILLANCE REQUIREMENTS

( continued)

PAM Instrumentation B 3.3.3.1 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. The Note is based upon a NRC Safety Evaluation Report (Ref. 2) which concluded that the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> testing allowance does not significantly reduce the probability of properly monitoring post accident parameters, when necessary.

SR 3.3.3.1.1 Performance of the CHANNEL CHECK once every 31 days ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel against a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; Alternately, the Frequency for checking the calibration of pressure, level, and flow transmitters may be determined in accordance with the Online Monitoring Program implemented in accordance with AMS-TR-0720R2-A (Ref. 4) and TS 5.5.17, Online Monitoring Program.

HATCH UNIT 2 The Surveillance Freq ency is controlled under the Surveillance Frequency Control Pro ram.

SR 3.3.3.1.2 CHANNEL CALI BRA TIO is a complete check of the instrument loop, including the sensor. The st verifies the channel responds to measured parameter with t necessary range and accuracy.

The Surveillance Frequenc controlled under the Surveillance Frequency Control Program.

B 3.3-68

( continued)

REVISION 79

PAM Instrumentation B 3.3.3.1 BASES (continued)

REFERENCES HATCH UNIT 2 Insert:

1.

Regulatory Guide 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Revision 2, December 1980.

2.

NRC Safety Evaluation Report, "Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, Conformance to Regulatory Guide 1.97,"

dated July 30, 1985.

3.

NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

4. AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters."

B 3.3-69 REVISION 79

BASES Remote Shutdown System B 3.3.3.2 SURVEILLANCE REQUIREMENTS SR 3.3.3.2.1 (continued)

REFERENCES The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.3.2.2 SR 3.3.3.2.2 verifies each required Remote Shutdown System transfer switch and control circuit performs the intended function. This verification is performed from the remote shutdown panel and locally, as appropriate. Operation of equipment from the remote shutdown Alternately, the Frequency for checking the calibration of pressure, level, and flow transmitters may be determined in accordance with the Online Monitoring Program implemented in accordance with AMS-TR-0720R2-A (Ref. 4) and TS 5.5.17, Online Monitoring Program.

SR 3.3.3.2.3 CHANNEL CALIBRATION is a omplete check of the instrument loop and the sensor. The test verifi the channel responds to measured parameter values with the nee ssary range and accuracy.

The Surveillance Frequenc i ontrolled under the Surveillance Frequency Control Program.

1.

10 CFR 50, Appendix A, GDC 19.

2.

Technical Requirements Manual, Table T6.0-1.

3.

NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

Insert:

4. AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters."

HATCH UNIT 2 B 3.3-74 REVISION 79

BASES SURVEILLANCE REQUIREMENTS Insert:

SR 3.3.4.2.3 (continued)

ATWS-RPT Instrumentation B 3.3.4.2 The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.4.2.4 The LOGIC SYSTEM FUN TIONAL TEST demonstrates the OPERABILITY of the requi ed trip logic for a specific channel. The system functional test oft e pump breakers is included as part of this Surveillance and overlap the LOGIC SYSTEM FUNCTIONAL TEST Alternately, the Frequency for checking the calibration of pressure, level, and flow transmitters may be determined in accordance with the Online Monitoring Program implemented in accordance with AMS-TR-0720R2-A (Ref. 4) and TS 5.5.17, Online Monitoring Program.

REFERENCES Insert:

1.

FSAR, Section 7.6.10.7.

2.

GENE-770-06-1, "Bases for Changes To Surveillance Test Intervals and Allowed Out-of-Service Times For Selected Instrumentation Technical Specifications," February 1991.

3.

NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

4. AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters."

HATCH UNIT 2 B 3.3-91 REVISION 79

BASES ECCS Instrumentation B 3.3.5.1 SURVEILLANCE REQUIREMENTS SR 3.3.5.1.1 (continued)

Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.

Alternately, the Frequency for checking the calibration of pressure, level, and flow transmitters may be determined in accordance with the Online Monitoring Program implemented in accordance with AMS-TR-0720R2-A (Ref. 10) and TS 5.5.17, Online Monitoring Program.

HATCH UNIT 2 The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.5.1.4 A CHANNEL CALIBRATION s a complete check of the instrument loop and the sensor. This te verifies the channel responds to the measured parameter within t necessary range and accuracy.

CHANNEL CALIBRATION le es the channel adjusted to account for instrument drifts between sue essive calibrations, consistent with the plant specific setpoint method logy.

The Surveillance Frequency I ontrolled under the Surveillance Frequency Control Program.

SR 3.3.5.1.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel.

The system functional testing performed in LCO 3.5.1, LCO 3.5.2, LCO 3.7.2, LCO 3.8.1, and LCO 3.8.2 overlaps this Surveillance to complete testing of the assumed safety function.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

B 3.3-122 (continued)

REVISION 134

BASES REFERENCES HATCH UNIT 2 Insert:

1.

FSAR, Section 5.2.

2.

FSAR, Section 6.3.

3.

FSAR, Chapter 15.

4.

(Not used)

ECCS Instrumentation B 3.3.5.1

5.

NEDC-30936-P-A, "BWR Owners' Group Technical Specification Improvement Analyses for ECCS Actuation Instrumentation, Part 2," December 1988.

6.

NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

7.

(Not used)

8.

(Not used)

9.

004N6160, "Edwin I. Hatch Nuclear Plant Units 1 and 2, TRACG-LOCA Loss-of-Coolant Accident Analysis," Revision 1, October 2019.

10. AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters."

(continued)

B 3.3-123 REVISION 134

BASES Primary Containment Isolation Instrumentation B 3.3.6.1 SURVEILLANCE REQUIREMENTS SR 3.3.6.1.1 (continued) instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.

Alternately, the Frequency for checking the calibration of pressure, level, and flow transmitters may be determined in accordance with the Online Monitoring Program implemented in accordance with AMS-TR-0720R2-A (Ref. 10) and TS 5.5.17, Online Monitoring Program.

HATCH UNIT 2 The Surveillance Frequency is c ntrolled under the Surveillance Frequency Control Program.

SR 3.3.6.1.3 SR 3.3.6.1.4 an SR 3.3.6.1.5 A CHANNEL CALIBRATION is loop and the sensor. This test measured parameter within the CHANNEL CALIBRATION leav instrument drifts between succ plant specific setpoint methodo The Surveillance Frequency i Frequency Control Program.

B3.3-168 complete check of the instrument rifies the channel responds to the ecessary range and accuracy.

s the channel adjusted to account for sive calibrations, consistent with the gy.

ntrolled under the Surveillance (continued)

REVISION 127

BASES SURVEILLANCE REQUIREMENTS Insert:

SR 3.3.6.1.6 Primary Containment Isolation Instrumentation B 3.3.6.1 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel.

The system functional testing performed on PCIVs in LCO 3.6.1.3 overlaps this Surveillance to provide complete testing of the assumed safety function. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.6.1. 7 This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis. The instrument response times must be added to the PCIV closure times to obtain the ISOLATION SYSTEM RESPONSE TIME.ISOLATION SYSTEM RESPONSE TIME acceptance criteria are included in Reference 6. This test may be performed in one measurement, or in overlapping segments, with verification that all components are tested.

A Note to the Surveillance states that channel sensors are excluded from ISOLATION SYSTEM RESPONSE TIME testing. The exclusion of the channel sensors is supported by Reference 8 which indicates that the sensors' response times are a small fraction of the total response time. Even if the sensors experienced response time degradation, they would be expected to respond in the microsecond to millisecond range until complete failure.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Alternately, the use of the allocated ISOLATION SYSTEM RESPONSE TIME for transmitters in the Online Monitoring Program is supported using the 'noise analysis' technique implemented in accordance with AMS-TR-0720R2-A (Ref. 10) to detect dynamic failures modes that can affect transmitter response time.

(continued)

HATCH UNIT 2 B3.3-169 REVISION 127

BASES REFERENCES Insert:

1.

2.

3.

4.

5.

6.

7.

8.

9.

Primary Containment Isolation Instrumentation B 3.3.6.1 FSAR, Section 6.3.

FSAR, Chapter 15.

FSAR, Paragraph 4.2.3.4.2.

NEDC-31677P-A, "Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation,"

July 1990.

NEDC-30851 P-A Supplement 2, "Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation," March 1989.

Technical Requirements Manual, Table T5.0-1.

NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

NEDO-32291, "System Analyses for Elimination of Selected Response Time Testing Requirements," January 1994.

004N6160, "Edwin I. Hatch Nuclear Plant Units 1 and 2, TRACG-LOCA Loss-of-Coolant Accident Analysis," Revision 1, October 2019.

10. AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters."

HATCH UNIT 2 B 3.3-170 REVISION 127

BASES SURVEILLANCE REQUIREMENTS REFERENCES Insert:

SR 3.5.1.12 (continued)

ECCS - Operating B 3.5.1 is performed by tests required by the ASME OM Code (Ref. 17).

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.5.1.13 This SR ensures that the ECCS RESPONSE TIMES are less than or equal to the maximum values assumed in the accident analysis.

Response time testing acceptance criteria are included in Reference 14. A Note to the Surveillance states that the instrumentation portion of the response time may be assumed from established limits. The exclusion of the instrumentation from the response time surveillance is supported by Reference 15, which concludes that instrumentation will continue to respond in the microsecond to millisecond range prior to complete failure.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program

1.

2.

3.

4.

Unit 2 FSAR, R ragraph 6.3.2.2.3.

,~r-----~L 6. AMS-TR-0720R2-A, "Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters."

Alternately, the use of the allocated ECCS RESPONSE TIME for transmitters in the Online Monitoring Program is supported using the 'noise analysis' technique implemented in accordance with AMS-TR-0720R2-A (Ref. 6) to detect dynamic failures modes that can affect transmitter response time.

(continued)

HATCH UNIT 2 B3.5-16 REVISION 130