AEP-NRC-2024-11, Unit 2 - Response to Request for Additional Information on Requested Change Regarding Neutron Flux Instrumentation

Unit 2 - Response to Request for Additional Information on Requested Change Regarding Neutron Flux Instrumentation

ML24058A357
Person / Time
Site:	Cook
Issue date:	02/27/2024
From:	Ferneau K Indiana Michigan Power Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
AEP-NRC-2024-11
Download: ML24058A357 (1)

Text

INDIANA MICHIGAN POWIR*

An MP Company BOUNDLESS ENERGY" Docket Nos.: 50-315 50-316 U. S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, D.C. 20555-0001 Donald C. Cook Nuclear Plant Unit 1 and Unit 2 Indiana Michigan Power Cook Nuclear Plant One Cook Place Bridgman, Ml 49106 indianamichiganpower.com AEP-NRC-2024-11 10 CFR 50.90 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON REQUESTED CHANGE REGARDING NEUTRON FLUX INSTRUMENTATION

References:

1. Letter from K. J. Ferneau, Indiana Michigan Power Company (l&M), to U.S. Nuclear Regulatory Commission (NRC),

"Request for Approval of Change Regarding Neutron Flux Instrumentation," dated January 26, 2023, Agencywide Documents Access and Management System (ADAMS) Accession No. ML23026A284.

2. Letter from Q. S. Lies, l&M, to NRC, "Supplement to Request for Approval of Change Regarding Neutron Flux Instrumentation," dated August 2, 2023, ADAMS Accession No. ML23214A289.

3. E-mail from S. P. Wall, NRC, to M. K. Scarpello, l&M, "Final RAI - D.C. Cook 1 & 2 - License Amendment Request Regarding Neutron Flux Instrumentation (EPID No. L-2023-LLA-0011 ),"

dated November 17, 2023, ADAMS Accession No. ML23321A122.

This letter provides Indiana Michigan Power Company's (l&M), licensee for Donald C. Cook Nuclear Plant (CNP) Unit 1 and Unit 2, response to the Request for Additional Information (RAI) by the U. S.

Nuclear Regulatory Commission (NRC) regarding a request to use alternate means of fulfilling the requirements of Regulatory Guide 1.97 with regards to the plant safety function of reactivity control.

The request would reclassify the wide range neutron flux instrumentation at CNP Unit 1 and Unit 2 as Category 3 instrumentation, and would modify Technical Specification (TS) Table 3.3.3-1, Post Accident Monitoring Instrumentation, to remove Function 1, Neutron Flux, from the list of required post-accident monitoring instrumentation. The existing TS require two channels of neutron flux instrumentation to be operable.

February 27, 2024

U.S Nuclear Regulatory Commission Page 2 AEP-NRC-2024-11 By Reference 1, l&M submitted a request for approval of the reclassification of the wide range neutron flux instrumentation to Category 3 and a corresponding change to CNP Unit 1 and Unit 2 TS Table 3.3.3-1, Post Accident Monitoring Instrumentation. By Reference 2, l&M submitted a supplement to Reference 1. By Reference 3, the NRC submitted an RAI concerning the letter submitted by l&M as Reference 1. to this letter provides an affirmation statement. Enclosure 2 to this letter provides l&M's response to the NRC's RAI from Reference 3.

As discussed with NRC staff during the public meeting held January 24, 2024 (ML24031A587),

related to this RAI response, l&M is requesting to revise the scope of the amendment request such that Neutron Flux remains as Function 1 in TS Table 3.3.3-1, Post Accident Monitoring Instrumentation, but is exempt from the requirement to maintain environmental qualification. A supplement to the amendment request, which addresses the scoping changes, is included as to this letter. and Enclosure 5 provide Unit 1 and Unit 2 TS pages, respectively, marked to show 'the proposed changes. and Enclosure 7 provide Unit 1 and Unit 2 TS Bases pages, respectively, marked to show the proposed changes. TS Bases markups are included for information only. Changes to the existing TS Bases, consistent with the technical and regulatory analysis, will be implemented under CNP's TS 5.5.12, "Technical Specifications Bases Control Program."

U.S Nuclear Regulatory Commission Page 3 AEP-NRC-2024-11 The changes proposed in this letter do not impact the conclusions provided in Reference 1 that a finding of "no significant hazards consideration" is justified. There are no new regulatory commitments made in this letter. Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Director, at (269) 466-2649.

Sincerely, Kelly J. Ferneau Site Vice President BMC/sjh

Enclosures:

1. Affirmation 2.

Response to Request for Additional Information on Requested Change Regarding Neutron Flux Instrumentation 3.

Supplement to License Amendment Request Regarding Neutron Flux Instrumentation 4.

Donald C. Cook Nuclear Plant Unit 1 Technical Specification Pages Marked to Show Proposed Changes 5.

Donald C. Cook Nuclear Plant Unit 2 Technical Specification Pages Marked to Show Proposed Changes 6.

Donald C. Cook Nuclear Plant Unit 1 Technical Specification Bases Pages Marked to Show Proposed Changes 7.

Donald C. Cook Nuclear Plant Unit 2 Technical Specification Bases Pages Marked to Show Proposed Changes c:

EGLE - RMD/RPS J. B. Giessner-NRC Region Ill NRC Resident Inspector N. Quilico - MPSC R. M. Sistevaris - AEP Ft. Wayne, w/o enclosures S. P. Wall-NRC Washington, D.C.

A. J. Williamson - AEP Ft. Wayne, w/o enclosures to AEP-NRC-2024-11 AFFIRMATION I, Kelly J. Ferneau, being duly sworn, state that I am the Site Vice President of Indiana Michigan Power Company (l&M), that I am authorized to sign and file this request with the U. S. Nuclear Regulatory Commission on behalf of l&M, and that the statements made and the matters set forth herein pertaining to l&M are true and correct to the best of my knowledge, information, and belief.

Indiana Michigan Power Company Kelly J. Ferneau Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS c97 DAY OF kbrua.r"I 2024

~~~~

My Commission Expires 0\\ I&\\ jac@o

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to AEP-NRC-2024-11 Response to Request for Additional Information on Requested Change Regarding Neutron Flux Instrumentation By letter dated January 26, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23026A284) (Reference 1 ), Indiana Michigan Power Company (l&M),

the licensee for Donald C. Cook Nuclear Plant (CNP) Unit 1 and Unit 2, submitted a request to use alternate means of fulfilling the requirements of Regulatory Guide (RG) 1.97 with regards to the plant safety function of reactivity control at CNP Unit 1 and Unit 2. The request would reclassify the wide range neutron flux instrumentation at CNP Unit 1 and Unit 2 as Category 3 instrumentation, and would modify Technical Specification (TS) Table 3.3.3-1, Post Accident Monitoring Instrumentation, to remove Function 1, Neutron Flux, from the list of required post-accident monitoring (PAM) instrumentation.

By letter dated August 2, 2023 (ADAMS Accession No. ML23214A289) (Reference 2), l&M submitted a supplement to Reference 1.

The U. S. Nuclear Regulatory Commission (NRC) staff is currently reviewing the submittal and has determined that additional information is needed in order to complete the review (Reference 3). The request for additional information (RAI) and l&M's response are provided below.

As discussed with NRC staff during the public meeting held January 24, 2024, related to this RAI response, l&M is requesting to revise the scope of the amendment request such that Neutron Flux remains as Function 1 in TS Table 3.3.3-1, Post Accident Monitoring Instrumentation, but is exempt from the requirement to maintain environmental qualification. A supplement to the amendment request, which addresses the scoping changes, is included as Enclosure 3 to this letter.

EICB-RAl-1 In the LAR, the licensee states, in part, that:

In a design basis accident, wide range neutron flux instrumentation provides information to control room operators in two situations - to check if the reactor is no longer critical and to monitor the core for unexpected additions of reactivity after reactor shutdown has been achieved.

During an accident that involves normal containment conditions, the neutron flux monitoring instrumentation is expected to be available for use. Under these conditions, control room operators are able to monitor the reactivity state of the core by evaluating neutron flux behavior measured by all of the neutron flux monitoring instrumentation (Gamma-Metrics in addition to Westinghouse power range, intermediate range, and source range Westinghouse instruments) as well as Core Exit Thermocouple (GET) temperatures, Reactor Coolant System (RCS) hot leg and cold leg temperatures, and boron concentration.

..... Additionally, the shutdown margin would be verified by measuring boron concentration.

The EOPs would also direct the operators to assure that boric acid injection is taking place, adding negative reactivity to ensure that the core remains shut down.

to AEP-NRC-2024-11 Page 2 CNP Unit 1 and Unit 2 EOPs also require control room operators to monitor RCS temperature using CETs and RCS hot leg and cold leg instruments, and to monitor boron concentration and the assurance of boron injection.

• Please provide information regarding how the indication of "boron concentration and the assurance of boron injection" instrumentation or sampling process may each be considered as key variables and be measured with high-quality instrumentation, in lieu of neutron flux monitoring (source range).

o Describe the response time characteristics of instrumentation used for responding to a change in boron concentration and for assuring boron injection is taking place.

o Describe whether the use of instrumentation measuring boron concentration will enable plant operators to take timely mitigative action in an event of a return to criticality following a LOCA event.

l&M Response to EICB-RAl-1 Boron concentration would not be considered as a key variable with regards to subcriticality, but would be used by control room operators to assess shutdown margin as part of an aggregate indication review. Continuous indication of boron concentration via instrumentation is not available at CNP Unit 1 or Unit 2, rather RCS boron concentration is measured by sampling the RCS.

While Emergency Core Cooling Flow is a Type A, Category 1 variable included in CNP Unit 1 and Unit 2 TS Table 3.3.3-1, Post Accident Monitoring Instrumentation, as Function 20, it is not considered a key variable with regards to subcriticality.

The Emergency Core Cooling System (ECCS) is designed to inject borated water from a combination of the Accumulators and the Refueling Water Storage Tank (RWST) to ensure that an adequate supply of borated water is added to the reactor vessel following a design basis accident. The design ensures boron injection through at least three intact loops with the entire contents of one loop conservatively assumed to be unavailable due to a break. The safety analysis and rigorous testing ensure that the injection that occurs through the three intact loops is adequate for all design basis accidents.

TS Table 3.3.3-1, Post Accident Monitoring Instrumentation, Function 20 prescribes the minimum equipment that is required for operability of an ECCS flow channel, and this is clarified by a note stating "Any combination of two instruments per train, including Centrifugal Charging Pump Flow, Safety Injection Pump Flow, Centrifugal Charging Pump Circuit Breaker Status, and Safety Injection Pump Circuit Breaker Status, can be used to satisfy Function 20 OPERABILITY requirements."

Step 7 of Emergency Operating Procedure OHP-4023-E-0, Reactor Trip or Safety Injection, directs an operator to perform Attachment A of the procedure which systematically reviews expected equipment response to start pumps and align equipment that is required for the accident conditions. This includes ensuring that ECCS pumps are operating with a flow path from the RWST to the RCS through their respective injection flow paths. This is confirmed by observing pump running currents, valve positions, and injection flow indication on the control panels.

to AEP-NRC-2024-11 Page3 EICB-RAl-2 Currently the Neutron Flux are fulfilling the requirements of 10 CFR 50.36(c)(2)(ii), as are the RCS Hot Leg Temperature (Wide Range), RCS Cold Leg Temperature (Wide Range), and GET monitoring instrumentations. If during an accident that involves elevated temperature containment conditions the neutron flux instrument is not available, the operators will be relying on boron concentration and the assurance of boron injection as key variables. Per RG 1.97 Revision 3, key variable should be qualified to meet Category design specifications. The CNP TS Bases states that the PAM instrumentation TSs ensures the operability of RG 1.97 Type A and Category 1 variables so that the control room operating staff can (among other items):

• Perform the diagnosis specified in the emergency operating procedures (these variables are restricted to preplanned actions for the primary success path of DBAs), e.g., LOCA.

• Take the specified, pre-planned manually controlled actions, for which no automatic control is provided, and that are required for safety systems to accomplish their safety functions.

• Determine whether systems important to safety are performing their intended functions.

Provide the basis for concluding that boron concentration monitoring instrumentations and the assurance of boron injection is not required to be added to the PAM table.

l&M Response to EICB-RAl-2 As discussed in response to EICB-RAl-1 above, boron concentration monitoring and the assurance of boron injection are not considered key variables with regards to subcriticality, though ECCS flow is a Type A, Category 1 variable and is included as Function 20 in CNP Unit 1 and Unit 2 TS Table 3.3.3-1, Post Accident Monitoring Instrumentation. Instrumentation to continuously monitor boron concentration is not available at CNP Unit 1 or Unit 2.

EICB-RAl-3 Currently the Neutron Flux monitoring instrumentations are Category 1, as are the RCS Hot Leg Temperature (Wide Range), RCS Cold Leg Temperature (Wide Range), and GET monitoring instrumentations. Describe how the boron concentration monitoring instrumentations meets the qualifications of Category 1. If not currently at Category 1 qualification, are there any plans to upgrading the qualification of boron concentration instrumentation to Category 1? Please describe if so.

l&M Response to EICB-RAl-3 As discussed in response to EICB-RAl-1 above, boron concentration is not considered a key variable with regards to subcriticality, but is considered a backup variable, as further discussed in response to EICB-RAl-4.

EICB-RAl-4 In the LAR, the licensee indicates that a DC Cook plant operator would rely on indications from core exit temperatures, hot and cold leg temperatures, boron concentration, and assurance of boron injection to verify that there is no continued or unexpected reactivity occurring. Any actions to be taken by the to AEP-NRC-2024-11 Page4 operator must be taken timely enough during the event to have a high degree of success in achieving or returning to cold shutdown, and hence a safe reactor state.

a. Demonstrate whether there will be adequate time for detection of process changes at the location of the CETs and whether there is appropriate instrument response time and sufficient time available from the onset of reactor shutdown, for an operator relying on CET or Hot/Cold Leg Temperature, and boron concentration indications to verify that reactor shutdown has been successfully accomplished through the insertion of the control rods or boron addition during an accident with energy added to the containment.

b. Demonstrate whether there will be adequate time for process changes at the location of the instruments and whether there is appropriate instrument response time and sufficient time available from the onset of unexpected reactivity, for an operator using GET or Hot/Cold Leg Temperature indications and boron concentration to observe that unexpected reactivity is occurring to enable timely action to mitigate this condition.

c.

Please provide an overview of an evaluation of expected process variations and the expected response of the GET and Hot/Cold Leg Temperature instruments to those variations regarding the time delay needed to allow for process changes to occur at the location of the Hot/Cold Leg Temperature instruments in response to those process variations.

Describe the expected response of these instruments to enable plant operators to take appropriate mitigative actions to recover from the accident and avoid further adverse consequences of the event.

Include your assumptions and conditions regarding whether the evaluation assumes whether the reactor coolant pumps are running and whether vessel or piping voiding conditions are occurring.

l&M Response to EICB-RAl-4 With regards to EICB-RAl-4 (a), as discussed with NRC staff during the public meeting held January 24, 2024, related to this RAI response, l&M is requesting to revise the scope of the amendment request such that Neutron Flux remains as Function 1 in TS Table 3.3.3-1, Post Accident Monitoring Instrumentation, but is exempt from the requirement to maintain environmental qualification.

For purposes of verifying initial reactor shutdown at CNP Unit 1 and Unit 2 Neutron Flux would continue to be the key variable. A supplement to the amendment request, which addresses the scoping changes, is included as Enclosure 3 to this letter.

The remainder of l&M's response to EICB-RAl-4 is intended to address items (b) and (c), regarding operator response to the onset of unexpected reactivity.

In all cases where adverse containment conditions would render the non-environmentally qualified source range detection instruments unreliable, the event will be accompanied by a Safety Injection signal.

Following completion of immediate actions, licensed operators prioritize starting and aligning any ECCS equipment that should have actuated automatically.

This is rapidly done using the operator's knowledge and will typically take place within the first two to three minutes of the reactor trip, depending on the operator's tasks for the event.

to AEP-NRC-2024-11 Page 5 Since ECCS flow is intended for both reactivity and inventory control, insertion of control rods, the build up of Xenon, and verification of ECCS injection ensure that significant negative reactivity is being added at the early stages of the accident without any reliance on nuclear instrumentation or Core Exit Thermocouple readings.

Steam Line Break Inside Containment While return to criticality and reactor power are not credible during the accident mitigation stage of a LOCA event, it is a possibility during the initial phase of a steam line break due to the large reactivity addition associated with an uncontrolled RCS cooldown. A steam line break inside containment could also create the conditions to render the non-environmentally qualified nuclear instrumentation unreliable. However, a postulated return to power for this type of cooldown event is self-limiting.

Assuming the initial reactor trip verification was successful, any return to criticality from an uncontrolled RCS cooldown during a steam line break event would be terminated through temperature feedback as the RCS heats up. The RCS temperature following the heat up would be below the temperature of the RCS at the time of the initial reactor trip, since boron would have been added by ECCS injection during the initial event response, and since control rods would insert during the reactor trip. The plant UFSAR accident analysis considers the return to power possibility from a steam line break, where the core is ultimately shut down by boric acid delivered by the ECCS to the RCS, which remains intact.

The recovery actions that follow include termination of ECCS injection, reestablishing normal charging and letdown, and eventually cooling down and depressurizing the RCS. ECCS termination is performed only after verifying procedural requirements for RCS inventory and subcooling are met. This ensures that Pressurizer Level is on scale and that no voids are present in the reactor vessel. At this point there is no accident in progress.

During post-accident recovery with the RCS intact, in a situation where Gamma-Metrics instruments are not available, control room operators are trained and directed by emergency operating procedures to monitor RCS temperature indication as a key variable to identify any postulated return to criticality and rising core power level. One or more indications of RCS temperature would be available to control room operators, including CET temperature, RCS Hot Leg temperature, and RCS Cold Leg temperature. These key parameters are all included in CNP Unit 1 and Unit 2 TS Table 3.3.3-1, Post Accident Monitoring Instrumentation, and conform to the design and qualification requirements of a Category 1 variable as described in RG 1.97, Revision 3 (Reference 4), other than deviations approved by the NRC in Reference 5.

If a natural circulation cooldown were required, procedural shutdown margin requirements are more stringent with no Reactor Coolant Pumps (RCPs) running to account for reduced mixing and longer loop transport times. Throughout the event, licensed operators and the Shift Technical Advisor verify that natural circulation cooling is occurring. Although loop transport times are slower without RCPs in service, temperature changes of just a few degrees Fahrenheit can be observed on RCS Hot Leg and Cold Leg temperature instruments and CETs.

In summary, the RCS temperature indication is responsive enough to be the key variable to monitor for a return to criticality when the RCS is intact.

In addition, Pressurizer Level with an intact RCS is very responsive to small changes in RCS temperature and would provide defense in depth for monitoring a return to criticality in this scenario.

While this parameter is in CNP Unit 1 and Unit 2 TS Table 3.3.3-1, Post Accident Monitoring Instrumentation, as Function 12, this parameter is considered a backup variable with respect to monitoring for a return to criticality.

to AEP-NRC-2024-11 Page6 Note that shutdown margin requirements are verified by RCS sampling prior to initiating an RCS cooldown, and the RCS is repeatedly sampled during the cooldown to ensure shutdown margin is maintained. However, RCS boron concentration sampling is considered a backup variable with respect to monitoring for a return to criticality.

Loss of Coolant Accident During the long-term recovery actions that follow a design basis LOCA event there are three credible dilution sources to consider. Essential Service Water (ESW) operational leakage through an out of service Containment Spray (CTS) heat exchanger, Component Cooling Water (CCW) leakage through an out of service Residual Heat Removal (RHR) heat exchanger, and Auxiliary Feedwater addition through a Steam Generator to a depressurized RCS are possible dilution sources. The remainder of services in and out of containment are isolated by either a Containment Isolation Phase A or Phase 8/CTS signal.

Dilution via CCW leakage is not considered a credible source of a return to criticality since the CCW surge tank level is trended by two independent level channels with an alarm function. Leak rates of less than 1 gpm are easily observable, allowing mitigating action to be taken before it could become a significant dilution source.

CNP Unit 1 and Unit 2 TS 3.4.13, Reactor Coolant System (RCS), limits Steam Generator tube leakage to 150 gallons per day; and, without assuming an additional failure, Auxiliary Feedwater operational leakage into the RCS would be bounded by the more credible case of ESW leakage through an out of service CTS heat exchanger.

Even with a relatively large ESW leakage rate postulated, the boron concentration dilution rate of the aggregate inventory recirculated through the reactor core, consisting of a mix of the initial reactor inventory, ECCS injection, and melted ice from the ice bed, would be relatively slow. If a return to criticality is postulated without identification by the plant operators, then the corresponding core power rise rate will also be relatively slow. In a situation where the Gamma-Metrics instruments are not available, control room operators are trained and directed by emergency operating procedures to monitor RCS temperature to identify a return to criticality and rising core power level. The key variables monitored are the CETs and Hot Leg and Cold Leg RCS temperature indications. These parameters are included in CNP Unit 1 and Unit 2 TS Table 3.3.3-1, Post Accident Monitoring Instrumentation, and conform to the design and qualification requirements of a Category 1 variable as described in RG 1.97, Revision 3 (Reference 4 ), other than deviations approved by the NRC in Reference 5.

Function Restoration Procedure OHP-4023-FR-S.1, Response to Nuclear Power Generation/A TWS, is exited when Wide Range Power is less than 5% and neutron flux is lowering. While the exact power level for fuel damage to occur would be dependent on many variables, including RCS pressure and reactor power distribution, reactor power less than 5% provides assurance that there is no imminent threat to a critical safety function from criticality. For the postulated dilution, the time to return to a core power level of 5% would be on a time scale of one to several hours. This slow rate of dilution would provide control room operators sufficient time to respond to a postulated return to criticality and rising core power level, even in the event that indications for CETs or Hot Leg and Cold Leg temperatures experience delays.

It is noted that a slow dilution of RCS boron concentration via ingress of unborated water would likely be identified by control room operators prior to returning to criticality using other variables. These other variables would be considered as backup variables with respect to subcriticality. In the case of a to AEP-NRC-2024-11 Page?

breached RCS, changes in Containment Water Level would be an effective indication of an in-progress dilution. The RCS boron concentration is also repeatedly sampled during post-accident conditions and would indicate if there was an in-progress dilution of the RCS.

Sample/Demonstration Calculation of Hypothetical Long-Term Dilution Rate A quantitative sample calculation was performed as a demonstration that the core dilution rate following a LOCA event will be relatively slow. It should be noted that the sample calculation was not designed or intended to be a bounding accident analysis case and was not intended to define any new plant design or licensing limits.

The sample calculation considers the effects of a 20 gallon per minute (gpm) leak through an out-of-service CTS heat exchanger diluting the recirculation sump inventory through the spray ring header following a LOCA event. Note that while there are no TS limits on CTS tube leakage, these heat exchangers are rigorously inspected by the Generic Letter 89-13 program, and a 20 gpm leak would be considered very large and detectable by plant personnel. As a point of reference, all four heat exchangers are currently below detectable values for leakage.

The sample calculation uses a simple dilution of solution chemistry formula (i.e., Lf=i Ci

• Vi = Cfinal

• Vtinat, where C and V are the concentrations and volumes, respectively, of the original and final solutions) to derive the amount of leakage volume in gallons to decrease the reactor vessel boron concentration. Key inputs and assumptions for the sample calculation include:

The aggregate solution volume of the recirculation inventory before any dilution effects includes the initial RCS volume, the volume introduced during safety injection, and the volume introduced by the melting of the ice bed.

The core is assumed to already be critical at zero power before any dilution of the aggregate solution occurs. This approach is conservative for this demonstration because a higher initial boron concentration maximizes the rate at which the dilution changes the boron concentration.

The boron concentration of the ice bed and the RWST for safety injection are conservatively assumed to be at the maximum concentration allowed by TS.

The boron concentration in the RCS is assumed to be relatively high, but not necessarily a bounding value.

The volume of water in the RWST is set to the TS minimum value, and it is assumed that operators transition the plant to recirculation mode prior to emptying the RWST, such that only about 69% of the volume of the RWST ends up inside containment.

The volume of melted ice from the containment ice bed corresponds to the TS minimum ice condenser ice mass value.

Based on the conservative assumptions described above, and a 20 gpm leak of unborated water, it is calculated that it would take nearly 300 gallons of water to dilute the recirculation inventory by 1 ppm, and the dilution would occur at roughly 4.1 ppm per hour. At this rate of boron dilution, the time duration to raise power from 0% to 5% on a hypothetical return to criticality would be on a time scale of one to several hours.

It is noted that this slow dilution rate would also provide plant operators significant time to diagnose and mitigate the dilution prior to a return to criticality occurring with the monitoring of backup variables such as post-accident RCS boron concentration sampling.

to AEP-NRC-2024-11 Page8 EICB-RAl-5 In the LAR, the licensee states, in part, that:

In addition, neutron flux instrumentation is not always proportional to reactor power, and therefore may provide anomalous indications which can potentially mislead the operator.

Excore neutron flux instrumentation response is dependent on the location of voi<;ling in the core and/or downcomer, the degree of core uncovery, and detector location. This is particularly likely for accidents which produce harsh containment environments since reactor vessel voiding may be occurring. Anomalous neutron flux indication (i.e., indication not proportional to reactor power) was observed at the Three Mile Island accident (Reference 3

[of LAR]) and has been demonstrated in NRC financed experiments (Reference 4 [of LAR]).

The NRC staff notes that one outcome of the Three Mile Island (TM/) recommendations was to have all PWR plants install a reactor vessel level indication system (RVLIS) to detect and monitor recovery from inadequate core cooling (ICC). All PWR plants were required to have redundant, environmentally qualified Class IE ICC systems. These systems were required to be functional during and following LOCA events. As described in Summary Report, "Westinghouse Reactor Vessel Level Instrumentation System for Monitoring Inadequate Core Cooling" dated December 1980 (ML181398695), Westinghouse-designed RVLIS systems are capable of monitoring reactor vessel upper head and plenum, and wide range (dynamic) level. Indication from this level system should be capable of informing plant operators of the location of any voiding occurring inside reactor vessel.

Information regarding the location of voiding in the reactor vessel should serve to support interpretation of readings from the neutron detectors outside reactor vessel which appear to be anomalous.

• Please describe the RVLIS installed at CNP and confirm that it meets the TM/ recommendations for monitoring ICC following a LOCA event.

• Please provide a description of the process or procedure the plant operators would use to interpret potentially anomalous neutron source range readings by using information from the RVLIS regarding the location of potential reactor vessel voiding.

l&M Response to EICB-RAl-5 At CNP Unit 1 and Unit 2 a RVLIS is provided to indicate the relative vessel water level or the relative void content of fluid in the vessel during post-accident conditions. This level indication assists the operator in recognizing conditions which may lead to high temperatures that could damage the vessel or its internals. Level indicators and recorders are located in the CNP Unit 1 and Unit 2 control rooms.

Sensors measuring the differential pressure, between the vessel head and the bottom and between the head and the hot legs, provide the basis for level indication. Because flow through the vessel affects differential pressure measurement, three level indication ranges are provided by separate sensors. One range monitors void content in the reactor vessel when one or more Reactor Coolant Pumps are running. The remaining two ranges monitor the entire vessel level and partial water level (from the top of the reactor head to the hot leg) at zero forced flow conditions (no Reactor Coolant Pump operating).

The differential pressure measurements are compensated for process effects using reactor coolant system pressure and temperature measurements. They are also compensated for environmental to AEP-NRC-2024-11 Page9 temperature effects on the RVLIS sensing lines using temperature measurements at representative sensing line locations.

At CNP Unit 1 and Unit 2 the RVLIS, also called the Reactor Coolant Inventory Tracking System, is included in TS Table 3.3.3-1, Post Accident Monitoring Instrumentation, as Function 6. The installed RVLIS conforms to the design and qualification requirements of a Category 1 variable per RG 1.97, Revision 3 (Reference 4 ), other than deviations approved by the NRC in Reference 5, and meets the TMI recommendations for monitoring for inadequate core cooling following a LOCA event.

At CNP Unit 1 and Unit 2 Function Restoration Procedure OHP-4023-FR-l-3, Response to Voids in Reactor Vessel, provides guidance to interpret RVLIS indication and take appropriate action. While this does not specifically address interpretation of source range indication, licensed operators and the Shift Technical Advisor are trained on the limitation of excore neutron instrumentation and would therefore expect anomalous indication any time subcooling requirements are not met.

References:

1. Letter from K. J. Ferneau, Indiana Michigan Power Company (l&M), to U.S. Nuclear Regulatory Commission (NRC),

"Request for Approval of Change Regarding Neutron Flux Instrumentation," dated January 26, 2023, Agencywide Documents Access and Management System (ADAMS) Accession No. ML23026A284.

2. Letter from Q. S. Lies, l&M, to NRC, "Supplement to Request for Approval of Change Regarding Neutron Flux Instrumentation," dated August 2, 2023, ADAMS Accession No. ML23214A289.

3. E-mail from S. P. Wall, NRC, to M. K. Scarpello, l&M, "Final RAI - D.C. Cook 1 & 2 - License Amendment Request Regarding Neutron Flux Instrumentation (EPID No. L-2023-LLA-0011 ),"

dated November 17, 2023, ADAMS Accession No. ML23321A122.

4. Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Revision 3, May 1983, ADAMS Accession No. ML003740282.

5. NRC letter, T. G. Colburn (NRC) to M. P. Alexich (Indiana Michigan Power Company),

"Emergency Response Capability - Conformance to Regulatory Guide 1.97 Revision 3 for the D. C. Cook Nuclear Plant, Units 1 and 2," dated December 14, 1990, ADAMS Accession No. ML17328A824.

to AEP-NRC-2024-11 Supplement to License Amendment Request Regarding Neutron Flux Instrumentation By letter dated January 26, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23026A284) (Reference 1 ), Indiana Michigan Power Company (l&M),

the licensee for Donald C. Cook Nuclear Plant (CNP) Unit 1 and Unit 2, submitted a request to use alternate means of fulfilling the requirements of Regulatory Guide (RG) 1.97 with regards to the plant safety function of reactivity control at CNP Unit 1 and Unit 2. The request would reclassify the wide range neutron flux instrumentation at CNP Unit 1 and Unit 2 as Category 3 instrumentation, and would modify Technical Specification (TS) Table 3.3.3-1, Post Accident Monitoring Instrumentation, to remove Function 1, Neutron Flux, from the list of required post-accident monitoring (PAM) instrumentation.

By letter dated August 2, 2023 (ADAMS Accession No. ML23214A289) (Reference 2), l&M submitted a supplement to Reference 1.

By Reference 3, the NRC submitted a Request for Additional Information (RAI) concerning the letter submitted by l&M as Reference 1.

Proposed Changes to License Amendment Request As discussed with NRC staff during the public meeting held January 24, 2024, l&M is requesting to revise the scope of the amendment request such that Neutron Flux remains as Function 1 in TS Table 3.3.3-1, Post Accident Monitoring Instrumentation, but a note would be added to the table indicating that Function 1 is exempt from the requirement to maintain environmental qualification (EQ). This change in scope is to address the inability to environmentally qualify one of the existing instruments at CNP Unit 2 due to lack of available parts and vendor support.

Background

In December of 1980, the NRC issued Revision 2 of Regulatory Guide (RG) 1.97, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident (Reference 4). This was followed by Revision 3 of RG 1.97 in May of 1983 (Reference 5). The stated purpose of RG 1.97 is to describe a method acceptable to the NRC staff for complying with the Commission's regulations to provide instrumentation to monitor plant variables and systems during and following an accident in a light-water-cooled nuclear power plant. Included in RG 1.97 Revision 2 and Revision 3 was the establishment of Neutron Flux as a Type B, Category 1 variable associated with the plant safety function of reactivity control, provided for the purposes of function detection and accomplishment of mitigation.

As stated in Reference 1, at CNP Unit 1 and Unit 2, in a design basis accident wide range neutron flux instrumentation provides information to control room operators in two situations - to verify the initial shutdown of the reactor following the accident and to monitor the core for unexpected additions of reactivity after reactor shutdown has been achieved.

Technical Analysis As described in Reference 1, for events that do not result in a harsh environment inside the containment volume, wide range neutron flux would still be available for use, regardless of whether to AEP-NRC-2024-11 Page 2 the instruments meet EQ criteria. When considering events that do result in an adverse containment atmosphere, Reference 1 describes how wide range neutron flux instrumentation is expected to be able to continue to be available to verify initial reactor shutdown, since the associated step in the emergency operating procedures will be performed before the neutron flux instrumentation is adversely impacted by the degrading conditions within the containment volume.

With regards to monitoring for unexpected additions of reactivity following an event that results in adverse containment conditions, Reference 1, Reference 2, and Enclosure 2 to this letter describe how control room operators would rely on information from the Core Exit Thermocouples (CETs) and the Reactor Coolant System (RCS) Hot and Cold Leg temperature instruments to identify an increase in RCS temperature associated with a return to criticality, and would be able to take timely action to mitigate this condition.

Regulatory Assessment RG 1.97, Revision 3 established Neutron Flux as a Type B, Category 1 variable with regards to the plant safety function of reactivity control, provided for the purposes of function detection and accomplishment of mitigation.

RG 1.97 states, in part, that a key variable is that single variable (or minimum number of variables) that most directly indicates the accomplishment of a safety function. It also states that the design and qualification criteria category assigned to each variable indicates whether the variable is considered to be a key variable or for system status indication or for backup or diagnosis, i.e., for Types B and C, the key variables are Category 1; backup variables are generally Category 3.

For the overall safety function of reactivity control, l&M considers the key variables to be Neutron Flux, CET temperature, and RCS Hot and Cold Leg temperatures. For purposes of verifying initial reactor shutdown l&M considers Neutron Flux to be the key variable. For events that do not result in an adverse containment environment Neutron Flux would still be available for use to monitor subcriticality following the initial reactor shutdown. For events that result in an adverse containment environment, l&M considers CET temperature and RCS Hot and Cold Leg temperatures to be the key variables used to monitor subcriticality following initial reactor shutdown.

In August 2016 the NRC issued Revision 6 of NUREG-0800, Standard Review Plan, Branch Technical Position 7-10, Guidance on Application of Regulatory Guide 1.97 (Reference 6). The stated objectives of this Branch Technical Position are to clarify the staff position on accident monitoring instrumentation, and to identify alternatives acceptable to the staff for satisfying the guidelines identified in RG 1.97.

With regards to the variable of Neutron Flux, Reference 6 Table 2, For PWRs: Acceptable Deviations and Clarifications to Revision 2 and 3 of Regulatory Guide 1.97, states that a non-environmentally qualified instrument is acceptable if qualified CETs and RCS Hot and Cold Leg temperature indications are provided in conjunction with directions in emergency procedures for operator action to assure that boric acid injection is occurring.

At CNP Unit 1 and Unit 2 CETs and RCS Hot and Cold Leg temperature instruments are Type A, Category 1 variables included in TS Table 3.3.3-1, Post Accident Monitoring Instrumentation. In addition, emergency operating procedures at CNP Unit 1 and Unit 2 direct operators to verify that boric acid injection is occurring. Thus the request to exempt Neutron Flux from the requirement to maintain environmental qualification is within the guidance provided by Reference 6 to AEP-NRC-2024-11 Page 3 Revisions to Text of Original License Amendment Request Due to the requested change in scope of the license amendment request, l&M proposes that the following substitutions be made to the text of Enclosure 2 to Reference 1. It should be noted that the changes proposed in this supplement do not impact the conclusions provided in Reference 1 that a finding of "no significant hazards consideration" is justified.

Update 1: Section 1.0, Summary Description, Paragraphs 1, 5, and 6 Original Paragraph 1 Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (l&M), the licensee for Donald C.

Cook Nuclear Plant (CNP) Unit 1 and Unit 2, is requesting U.S. Nuclear Regulatory Commission (NRC) approval to use alternate means of fulfilling the requirements of Regulatory Guide (RG) 1.97 with regards to the plant safety function of reactivity control. l&M is requesting to reclassify the wide range neutron flux instrumentation at CNP Unit 1 and Unit 2 as Category 3 instrumentation, and is requesting a corresponding change to the Technical Specification (TS) for CNP Unit 1 and Unit 2. l&M proposes to modify TS Table 3.3.3-1, Post Accident Monitoring Instrumentation, to remove Function 1, Neutron Flux, from the list of required post-accident monitoring (PAM) instrumentation. The existing TS require two channels of neutron flux instrumentation to be operable.

Revised Paragraph 1 Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (l&M), the licensee for Donald C.

Cook Nuclear Plant (CNP) Unit 1 and Unit 2, is requesting U.S. Nuclear Regulatory Commission (NRC) approval to add a note to CNP Unit 1 and Unit 2 TS Table 3.3.3-1, Post Accident Monitoring Instrumentation, Function 1, Neutron Flux, exempting channels used to satisfy Function 1 from the requirement to be environmentally qualified. The existing TS require the two channels of neutron flux instrumentation to be environmentally qualified in order to be considered operable.

Original Paragraph 5 and 6 The proposed change to the CNP Unit 1 and Unit 2 TS would remove neutron flux from the list of required PAM instrumentation because other existing instrumentation remains available to control room operators to confirm that the reactor is no longer critical.

The proposed reclassification of wide range neutron flux instrumentation, and associated TS change, would more closely align with the role of the wide range neutron flux instrumentation as backup instrumentation for the purposes of PAM at CNP Unit 1 and Unit 2 and would allow l&M to pursue resolution of an inoperable neutron flux channel without undue risk of a TS-required shut down.

Revised Paragraph 5 and 6 The proposed change to the CNP Unit 1 and Unit 2 TS would allow non-environmentally qualified neutron flux instruments to satisfy the requirements of TS Table 3.3.3-1 Function 1, Neutron Flux, because environmental qualification is not required in order for the neutron flux instrumentation to accomplish its role as the key variable used to confirm initial reactor shutdown following a reactor trip or safety injection.

For events that do not result in an adverse containment to AEP-NRC-2024-11 Page4 environment Neutron Flux would continue to be available for use to monitor subcriticality following the initial reactor shutdown. For events that result in an adverse containment environment, l&M considers CET temperature and RCS Hot and Cold Leg temperatures to be the key variables used to monitor subcriticality following initial reactor shutdown.

The proposed exemption of environmental qualification requirements for TS-required neutron flux instrumentation would allow the continued use of the Gamma-Metrics neutron flux instruments for the purposes of post-accident monitoring at CNP Unit 1 and Unit 2 without adverse impact to the ability of control room operators to respond to an event.

Update 2: Section 2.3, Reason for the Proposed Change, Paragraph 5 Original Paragraph Reclassifying the wide range neutron flux instrumentation at CNP Unit 1 and Unit 2 as Category 3 instrumentation, and removing the instrumentation from TS Table 3.3.3-1 would allow l&M to pursue resolution of an inoperable wide range neutron flux instrument without undue risk of a TS-required shutdown. This becomes particularly important as the instrument vendor prepares to end support an~ maintenance of the currently installed instrumentation.

Revised Paragraph The proposed exemption of environmental qualification requirements for TS-required neutron flux instrumentation would allow the continued use of the Gamma-Metrics neutron flux instruments for the purposes of post-accident monitoring at CNP Unit 1 and Unit 2 without adverse impact to the ability of control room operators to respond to an event. This becomes particularly important as the instrument vendor prepares to end support and maintenance of the currently installed instrumentation, and support of the instruments transitions to a new vendor, with timelines for parts availability still uncertain, impacting l&M's ability to environmentally qualify one of the existing instruments at CNP Unit 2.

Update 3: Section 2.4, Description of the Proposed Change, Entire Section Updated Text (replaces existing section) l&M is requesting NRC approval to add a note to CNP Unit 1 and Unit 2 TS Table 3.3.3-1, Post Accident Monitoring Instrumentation, Function 1, Neutron Flux, exempting channels used to satisfy Function 1 from the requirement to be environmentally qualified.

The proposed change to CNP Unit 1 and Unit 2 TS would also supersede the regulatory commitment made by l&M in Reference 1 to provide two channels of neutron flux instrumentation that meet Category 1 requirements, including environmental qualification.

Two channels of neutron flux instrumentation would remain installed, but environmental qualification of the neutron flux instrumentation would no longer be required.

No changes are requested to CNP Unit 1 nor Unit 2 TS 3.9.2, Nuclear Instrumentation.

to AEP-NRC-2024-11 Page 5 Update 4: Section 4.1, Applicable Regulatory Requirements/Criteria, Paragraphs 6, 9 and 10 Original Paragraph 6 l&M proposes an alternative method of fulfilling the requirements of RG 1.97 with regards to reactivity control.

It is proposed that RCS Hot Leg Water Temperature, RCS Cold Leg Temperature, and Core Exit Temperature be considered as Type B variables for reactivity control, for the purpose of function detection, and that the associated instrumentation meet the requirements of Category 1 instrumentation.

Neutron flux would be considered as backup instrumentation for the purpose of verification of reactivity control, and a Category 3 classification is appropriate.

Revised Paragraph 6 l&M proposes an alternative method of fulfilling the requirements of RG 1.97 with regards to reactivity control.

It is proposed that RCS Hot Leg Water Temperature, RCS Cold Leg Temperature, and Core Exit Temperature be considered as key variables for the purpose of identifying unexpected reactivity following an accident that involves an adverse containment environment, and that the associated instrumentation meet the requirements of Category 1 instrumentation. Neutron flux would be considered as a key variable for the purpose of verifying initial reactor shutdown following an accident and would be available for events that do not result in an adverse containment environment, and would continue to meet the requirements of Category 1 instrumentation with the exception that environmental qualification would not be required.

Original Paragraphs 9 and 10 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants, requires licensee of nuclear plants to establish a program for qualifying certain electrical equipment important to safety, including environmental qualification. The equipment covered by this section includes safety-related equipment that is relied upon to remain functional during and following design basis events to ensure ( 1) the integrity of the reactor coolant pressure boundary, (2) the capability to shut down the reactor and maintain it in a safe shutdown condition, or (3) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite radiation exposures. Also covered by this section is certain post-accident monitoring equipment, per guidance provided in Revision 2 of RG 1.97. As discussed above, neutron flux would be considered as backup instrumentation for the purpose of verification of reactivity control, and while control room operators would continue to use the information provided by the Gamma-Metrics instruments as long as it is available, the instruments themselves would not be relied upon in the event of an adverse containment environment. Thus, with neutron flux considered as backup instrumentation, the requirements of 10 CFR 50.49 would not dictate that neutron flux instrumentation be environmentally qualified.

10 CFR 50.36(c)(2)(ii) provides four criteria that would necessitate establishing a TS limiting condition for operation. When considering neutron flux as a Category 3 variable, consistent with its use as backup instrumentation for PAM at CNP Unit 1 and Unit 2, none of the four criteria apply, and removal from TS Table 3.3.3-1, Post Accident Monitoring Instrumentation, is justified.

to AEP-NRC-2024-11 Page 6 Revised Paragraph 9 (Paragraph 10 would be deleted) 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants, requires licensee of nuclear plants to establish a program for qualifying certain electrical equipment important to safety, including environmental qualification. The equipment covered by this section includes safety-related equipment that is relied upon to remain functional during and following design basis events to ensure ( 1) the integrity of the reactor coolant pressure boundary, (2) the capability to shut down the reactor and maintain it in a safe shutdown condition, or (3) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite radiation exposures. Also covered by this section is certain post-accident monitoring equipment, per guidance provided in Revision 2 of RG 1.97. As discussed above, neutron flux would be considered as a key variable for verifying reactor shutdown following an accident, and while control room operators would continue to use the information provided by the Gamma-Metrics instruments as long as it is available, the instruments themselves would not be required to withstand the effects of an adverse containment environment in order to accomplish this function. Thus, while neutron flux instrumentation is expected to remain functional for a brief period following the start of a design basis accident, the requirements of 10 CFR 50.49 would not dictate that neutron flux instrumentation be environmentally qualified.

Update 5: Section 4.2, No Significant Hazards Determination, Entire Section Updated Text (replaces existing section)

Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (l&M), the licensee for Donald C.

Cook Nuclear Plant (CNP) Unit 1 and Unit 2, is requesting U.S. Nuclear Regulatory Commission (NRC) approval to add a note to CNP Unit 1 and Unit 2 TS Table 3.3.3-1, Post Accident Monitoring Instrumentation, Function 1, Neutron Flux, exempting channels used to satisfy Function 1 from the requirement to be environmentally qualified. The existing TS require the two channels of neutron flux instrumentation to be environmentally qualified in order to be considered operable.

l&M has evaluated whether or not a significant hazards consideration is involved with the proposed TS Bases change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No.

The proposed TS change involves elimination of environmental qualification (EQ) requirements from a specific set of required PAM instrumentation. The proposed change does not involve a physical change to the plant or a change in the manner in which the plant is operated or controlled.

The PAM instrumentation provides information to control room operators after an accident has occurred. Therefore, the probability of occurrence of an accident previously evaluated is not significantly increased by the proposed amendment.

The proposed change removes EQ requirements from a specific set of required PAM instrumentation, but ensures that the information required by control room operators is still available in the event of an accident, thus not significantly increasing the consequences of an accident previously evaluated.

to AEP-NRC-2024-11 Page 7 Therefore, it is concluded that the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed TS change removes EQ requirements from a specific set of required PAM instrumentation and does not alter the design function or operation of any structure, system, or component that may be involved in the initiation of an accident. The proposed change does not create new failure mechanisms, malfunctions, or accident initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed TS change involves elimination of EQ requirements from a specific set of required PAM instrumentation. This change does not alter the manner in which safety limits, limiting safety system setpoints, or limiting conditions for operation are determined. The EQ requirements of a specific set of required PAM instrumentation are changed, but the necessary information available to control room operators is retained.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

In conclusion, based on the considerations discussed above, (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 NRC's regulations, and (3) approval of the proposed TS change will not be inimical to the common defense and security or to the health and safety of the public. l&M concludes that the proposed TS change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.

References:

1. Letter from K. J. Ferneau, Indiana Michigan Power Company (l&M), to U.S. Nuclear Regulatory Commission (NRC),

"Request for Approval of Change Regarding Neutron Flux Instrumentation," dated January 26, 2023, Agencywide Documents Access and Management System (ADAMS) Accession No. ML23026A284.

2. Letter from Q. S. Lies, l&M, to NRC, "Supplement to Request for Approval of Change Regarding Neutron Flux Instrumentation," dated August 2, 2023, ADAMS Accession No. ML23214A289.

3. E-mail from S. P. Wall, NRC, to M. K. Scarpello, l&M, "Final RAI - D.C. Cook 1 & 2 - License Amendment Request Regarding Neutron Flux Instrumentation (EPID No. L-2023-LLA-0011 ),"

dated November 17, 2023, ADAMS Accession No. ML23321A122.

to AEP-NRC-2024-11 Page 8

4. 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, ADAMS Accession No. ML060750525.

5. Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Revision 3, May 1983, ADAMS Accession No. ML003740282.

6. NUREG-0800, Standard Review Plan -Chapter 7, Branch Technical Position 7-10, Revision 6, Guidance on Application of Regulatory Guide 1.97, dated August 2016, ADAMS Accession No. ML16019A169.

to AEP-NRC-2024-11 Donald C. Cook Nuclear Plant Unit 1 Technical Specification Pages Marked to Show Proposed Changes

Table 3.3.3-1 (page 1 of 2)

Post Accident Monitoring Instrumentation FUNCTION REQUIRED CHANNELS

1.

Neutron Flux 2~

2.

Steam Generator Pressure (per steam generator) 2

3.

Reactor Coolant System (RCS) Hot Leg 2

Temperature (Wide Range)

4.

RCS Cold Leg Temperature (Wide Range) 2

5.

RCS Pressure (Wide Range) 2

6.

Reactor Coolant lnventol)' Tracking System 2

(Reactor Vessel Level Indication)

7.

Containment Water Level 2~

8.

Containment Pressure (Narrow Range) 2

9.

Penetration Flow Path Containment Isolation Valve 2 per penetration flow Position path!l>l<cl~

10.

Containment Area Radiation (High Range) 2 11.

Deleted

12.

Pressurizer Level 2

13.

Steam Generator Water Level (Wide Range) 4

14.

Condensate Storage Tank Level

15.

Core Exit Temperature - Quadrant 1 2~

16.

Core Exit Temperature - Quadrant 2 2~

17.

Core Exit Temperature - Quadrant 3 2~

18.

Core Exit Temperature - Quadrant 4 2~

I (a) Channels used to satisfy Function 1 are not required to be environmentally qualified.

PAM Instrumentation 3.3.3 CONDITION REFERENCED FROM REQUIRED ACTION E.1 F

F F

F F

G F

F F

G F

F G

F F

F F

~

Up to one channel of Function 7 OPERABILITY requirements can be satisfied by an OPERABLE train of containment water level switches if both Containment Water Level channels are inoperable. This substitution is only allowed until the end of the current operating cycle when it is invoked.

~

Not required for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.

m Only one position indication channel is required for penetration flow paths with only one installed control room indication channel.

~

A channel consists of one core exit thennocouple (CET).

Cook Nuclear Plant Unit 1 3.3.3-4 Amendment No. ~. 343, 3eG

Table 3.3.3-1 (page 2 of 2)

Post Accident Monitoring Instrumentation FUNCTION REQUIRED CHANNELS

19.

Secondary Heat Sink Indication 2~

(per steam generator)

20.

Emergency Core Cooling System Flow (per train) 2~

21.

Containment Pressure (Wide Range) 2

22.

Refueling Water Storage Tank Level 2

23.

RCS Subcooling Margin Monitor 1~

24.

Component Cooling Water Pump Circuit Breaker 2

Status

25.

Containment Recirculation Sump Water Level 2

PAM Instrumentation 3.3.3 CONDITION REFERENCED FROM REQUIRED ACTION E.1 F

F F

F F

G F

~

Any combination of two instruments per steam generator, including Steam Generator Water Level (Narrow Range) and Auxiliary Feedwater Flow, can be used to satisfy Function 19 OPERABILITY requirements.

~

Any combination of two instruments per train, including Centrifugal Charging Pump Flow, Safety Injection Pump Flow, Centrifugal Charging Pump Circuit Breaker Status, and Safety Injection Pump Circuit Breaker Status, can be used to satisfy Function 20 OPERABILITY requirements.

~

An OPERABLE plant process computer (PPG) subcooling margin readout can be used as a substitute for an inoperable Function 23, RCS Subcooling Margin Monitor.

Cook Nuclear Plant Unit 1 3.3.3-5 Amendment No. ~. 299, ~. 3eQ to AEP-NRC-2024-11 Donald C. Cook Nuclear Plant Unit 2 Technical Specification Pages Marked to Show Proposed Changes

Table 3.3.3-1 (page 1 of 2)

Post Accident Monitoring Instrumentation FUNCTION REQUIRED CHANNELS

1.

Neutron Flux 2~

2.

Steam Generator Pressure (per steam generator) 2

3.

Reactor Coolant System (RCS) Hot Leg Temperature (Wide Range) 2

4.

RCS Cold Leg Temperature (Wide Range) 2

5.

RCS Pressure (Wide Range) 2

6.

Reactor Coolant Inventory Tracking System 2

(Reactor Vessel Level Indication)

7.

Containment Water Level 2~

8.

Containment Pressure (Narrow Range) 2

9.

Penetration Flow Path Containment Isolation Valve 2 per penetration flow Position pathl&llc~

10.

Containment Area Radiation (High Range) 2

11.

Deleted

12.

Pressurizer Level 2

13.

Steam Generator Water Level (Wide Range) 4

14.

Condensate Storage Tank Level 1

15.

Core Exit Temperature - Quadrant 1 2!d~

16.

Core Exit Temperature - Quadrant 2 2!d~

17.

Core Exit Temperature - Quadrant 3 2{<1~

18.

Core Exit Temperature - Quadrant 4 2!df:I I (a) Channels used to satisfy Function 1 are not required to be environmentally qualified.

PAM Instrumentation 3.3.3 CONDITION REFERENCED FROM REQUIRED ACTION E.1 F

F F

F F

G F

F F

G F

F G

F F

F F

~

Up to one channel of Function 7 OPERABILITY requirements can be satisfied by an OPERABLE train of containment water level switches if both Containment Water Level channels are inoperable. This substitution is only allowed until the end of the current operating cycle when it is invoked.

~

Not required for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.

~

Only one position indication channel is required for penetration flow paths with only one installed control room indication channel.

~

A channel consists of one core exit thermocouple (CET).

Cook Nuclear Plant Unit 2 3.3.3-4 Amendment No. ~. ~. ~

Table 3.3.3-1 (page 2 of 2)

Post Accident Monitoring Instrumentation FUNCTION REQUIRED CHANNELS

19.

Secondary Heat Sink Indication 2!erJ (per steam generator)

20.

Emergency Core Cooling System Flow (per train) 2{1)~

21.

Containment Pressure (Wide Range) 2

22.

Refueling Water Storage Tank Level 2

23.

RCS Subcooling Margin Monitor 119~

24.

Component Cooling Water Pump Circuit Breaker 2

Status

25.

Containment Recirculation Sump Water Level 2

PAM Instrumentation 3.3.3 CONDITION REFERENCED FROM REQUIRED ACTION E.1 F

F F

F F

G F

~

Any combination of two instruments per steam generator, including Steam Generator Water Level (Narrow Range) and Auxiliary Feedwater Flow, can be used to satisfy Function 19 OPERABILITY requirements.

~

Any combination of two instruments per train, including Centrifugal Charging Pump Flow, Safety Injection Pump Flow, Centrifugal Charging Pump Circuit Breaker Status, and Safety Injection Pump Circuit Breaker Status, can be used to satisfy Function 20 OPERABILITY requirements.

~

An OPERABLE plant process computer (PPC) subcooling margin readout can be used as a substitute for an inoperable Function 23, RCS Subcooling Margin Monitor.

Cook Nuclear Plant Unit 2 3.3.3-5 Amendment No. 2e9, ~. ~. ~

to AEP-NRC-2024-11 Donald C. Cook Nuclear Plant Unit 1 Technical Specification Bases Pages Marked to Show Proposed Changes

PAM Instrumentation B 3.3.3 B 3.3 INSTRUMENTATION B 3.3.3 Post Accident Monitoring (PAM) Instrumentation BASES BACKGROUND APPLICABLE SAFETY ANALYSES 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 (DBAs).

The OPERABILITY of the accident monitoring instrumentation ensures that there is sufficient information available on selected unit parameters to monitor and to assess unit status and behavior following an accident.

The availability of accident monitoring instrumentation is important so that responses to corrective actions can be observed and the need for, and magnitude of, further actions can be determined. These essential instruments are identified in References 1, 2,ffi;J and al] addressing the recommendations of Regulatory Guide 1.97 (Ref. 3) as required by Supplement 1 to NUREG-0737 (Ref. 4).

The instrument channels required to be OPERABLE by this LCO include two classes of parameters identified during unit specific implementation of Regulatory Guide 1.97 as Type A and Category 1 variables.

These key variables are identified by the unit specific Regulatory Guide 1.97 analyses (Ref. 1, 2,ffi;J and e[zj). These analyses identify the unit specific Type A and Category 1 variables and provide justification for deviating from the NRC guidance in Reference 3.

The specific instrument Functions listed in Table 3.3.3-1 are discussed in the LCO section.

The PAM instrumentation LCO ensures the OPERABILITY of Regulatory Guide 1.97 Type A variables so that the control room operating staff can:

Perform the diagnosis specified in the emergency operating procedures (these variables are restricted to preplanned actions for the primary success path of DBAs), e.g., loss of coolant accident (LOCA); and Take the specified, pre-planned, manually controlled actions, for which no automatic control is provided, and that are required for safety systems to accomplish their safety function.

Cook Nuclear Plant Unit 1 B 3.3.3-1 Revision No. 48

BASES LCO (continued)

PAM Instrumentation B 3.3.3 One exception to the two channel requirement is Containment Isolation Valve (CIV) Position. In this case, the important information is the status of the containment penetrations. The LCO requires one position indicator for each active CIV. This is sufficient to redundantly verify the isolation status of each isolable penetration either via indicated status of the active valve and prior knowledge of a passive valve, or via system boundary status. If a normally active CIV is known to be closed and deactivated, position indication is not needed to determine status. Therefore, the position indication for valves in this state is not required to be OPERABLE.

Type A and Category 1 variables meet Regulatory Guide 1.97 Category 1 (Ref. 3) design and qualification requirements for seismic and environmental qualification, single failure criterion, utilization of emergency standby power, immediately accessible display, continuous readout, and recording of ~lay, except for approved deviations, as described in References 1l11 and ~

Listed below are discussions of the specified instrument Functions listed in Table 3.3.3-1. For all applicable Functions, the recorder or indicator may be used as the qualified instrument.

1.

Neutron Flux Neutron Flux (NRl-21 and NRl-23) is a Category 1 variable provided to verify reactor shutdown. The range of each of the two neutron flux instruments ( 1 0E-8 to 200% power) covers the full range of flux that may occur post accident.

As stated in Note (a) to Table 3.3.3-1, neutron flux instruments are not required to be environmentally qualified to be considered OPERABLE. This is acceptable because verification of initial reactor shutdown is expected to be completed prior to any potential impact to the neutron flux instrumentation due to an adverse containment environment. Other instruments will be used to monitor for subcriticality in the event of an adverse containment environment.

2.

Steam Generator (SG) Pressure (per SG)

Steam Generator Pressure is a Type A, Category 1 variable provided for determination of required core exit temperature. Three steam generator pressure channels per steam generator are provided (MPP-210, MPP-211, MPP-212, MPP-220, MPP-221, MPP-222, MPP-230, MPP-231, MPP-232, MPP-240, MPP-241, and MPP-242).

Each channel has a range of 0 psig to 1200 psig. However, only two steam generator pressure channels per steam generator are required Cook Nuclear Plant Unit 1 83.3.3-3 Revision No. 44

BASES LCO (continued)

PAM Instrumentation B3.3.3 3, 4.

Reactor. Coolant System (RCS) Hot and Cold Leg Temperatures (Wide Range)

RCS Hot and Cold Leg Temperatures are Type A, Category 1 variables provided for verification of core cooling and long term surveillance. RCS hot and cold leg tern eratures are used to determine RCS subcooling margin nd to monitor for subcriticalit in he event of an adverse containment environmen.

The RCS hot leg and RCS cold leg channels each receive input from one resistance temperature detector (RTD). In each of RCS loops 1 and 3, there is one RCS hot leg RTD (NTR-110 with MR-9, and NTR-130 with MR-11) and one RCS cold leg RTD (NTR-210 with MR-9, and NTR-230 with MR-11) that satisfy the guidance of Reference 3. The channels provide indication over a range of 0°F to 700°F.

5.

RCS Pressure (Wide Range)

RCS wide range pressure is a Type A, Category 1 variable provided for verification of core cooling and RCS integrity long term surveillance.

RCS wide range pressure is used as criteria to manually trip the reactor coolant pumps.

In addition, RCS wide range pressure is used for determining RCS subcooling margin.

Two RCS Pressure (Wide Range) channels are provided (NPS-110 and NPS-111, with MR-13), each with a range of O psig to 3000 psig.

6.

Reactor Coolant Inventory Tracking System {Reactor Vessel Level Indication)

Reactor coolant inventory is a Category 1 variable provided for verification and long term surveillance of core cooling.

The Reactor Coolant Inventory Tracking System consists of two channels of instrumentation (NLl-110, NLl-111, NLl-120, NLl-121, NLl-130, and NLl-131). Each channel is capable of measuring upper plenum level, narrow range level, and dynamic head (i.e., wide range level). The Reactor Coolant Inventory Tracking System provides a direct measurement of the collapsed liquid level above the fuel alignment plate. The collapsed level represents the amount of liquid mass that is in the reactor vessel above the core. Measurement of Cook Nuclear Plant Unit 1 B 3.3.3-4 Revision No. 0

BASES LCO (continued)

13.

Steam Generator Water Level (Wide Range)

PAM Instrumentation B 3.3.3 SG Water Level is a Category 1 variable provided to monitor operation of decay heat removal via the SGs. Four steam generator level (wide range) channels (one per steam generator) are provided (BLl-110, BLl-120, BLl-130, and BLl-140). Each channel is capable of monitoring from 12 inches above the steam generator tube sheet to the separators.

14.

Condensate Storage Tank (CST) Level CST Level is a Category 1 variable provided to ensure water supply for auxiliary feedwater (AFW). The CST provides the qualified water supply for the AFW System. Inventory is monitored from essentially the top of the CST to the bottom of the CST (95% total volume) by a single channel provided to satisfy the guidance of Reference 3, as described in Reference 1. CST Level is displayed on a control room indicator (CLl-114).

15, 16, 17, 18.

Core Exit Temperature Core Exit Temperature is a Type A, Category 1 variable used to determine whether to manually reduce ECCS flow. This variable is also provided for verification and long term surveillance of core cooling. In addition, core exit tern erature is used for determinin RCS subcooling margin nd to monitor for subcriticalit in the even f an adverse containment environmen.

Two OPERABLE channels of Core Exit Temperature, with one core exit thermocouple per channel, are required in each quadrant to provide indication of radial distribution of the coolant temperature rise across representative regions of the core. Two core exit temperature channels per quadrant ensure a single failure will not disable the ability to determine the radial temperature gradient. Each core exit temperature channel (SG-30 and SG-31 for TC 1 through 65) has a range of 200°F to 2300°F.

19.

Secondary Heat Sink Indication (per SG)

Cook Nuclear Plant Unit 1 Secondary Heat Sink Indication is a Type A, Category 1 variable used to determine whether to manually reduce ECCS flow. This variable is also provided to monitor operation of decay heat removal via the SGs.

As stated in Note ~

to Table 3.3.3-1, the requirements for this variable are met by any combination of two instruments per SG, B 3.3.3-7 Revision No. Q

BASES PAM Instrumentation B 3.3.3 SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.3.2 Deleted REFERENCES SR 3.3.3.3 CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter with the necessary range and accuracy. This SR is modified by a Note that excludes neutron detectors. For Function 9, the CHANNEL CALIBRATION shall consist of verifying that the position indication conforms to actual valve position. For Functions 15, 16, 17, and 18, whenever a sensing element is replaced, the next required CHANNEL CALIBRATION of the Core Exit Temperature thermocouple sensors is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing elements.

For Functions 20 (Circuit Breaker Status channels) and 24, the CHANNEL CALIBRATION shall consist of verifying that the position indication conforms to actual circuit breaker position. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

1.

NRC letter, T. G. Colburn (NRC) to M. P. Alexich (Indiana Michigan Power Company), "Emergency Response Capability-Conformance to Regulatory Guide 1.97 Revision 3 for the D. C. Cook Nuclear Plant, Units 1 and 2," dated December 14, 1990.

2.

UFSAR, Table 7.8-1.

3.

Regulatory Guide 1.97, Revision 3, May 1983.

4. NUREG-0737, Supplement 1, "TMI Action Items."

5. NRC letter, P. S. Tam (NRC), to M. K. Nazar, (Indiana Michigan Power Company), "Donald C. Cook Nuclear Plant, Units 1 & 2 (DCCNP-1 AND DCCNP-2) - Issuance of Amendments Re: Containment Sump Modifications per Generic Letter 2004-02 (TAC Nos. MD5901 AND MD5902)," dated October 18, 2007.

6. Letter from Indiana Michigan Power Company (K. J. Ferneau) to the NRC dated XXXX, XX 2024.

7. NRC letter, X. X. XXXXXX (NRC), to Q. S. Lies (Indiana Michigan Power Company), XXXX, dated XXXX XX, 2024.

Cook Nuclear Plant Unit 1 B 3.3.3-14 Revision No. 93 to AEP-NRC-2024-11 Donald C. Cook Nuclear Plant Unit 2 Technical Specification Bases Pages Marked to Show Proposed Changes

PAM Instrumentation B 3.3.3 B 3.3 INSTRUMENTATION B 3.3.3 Post Accident Monitoring (PAM) Instrumentation BASES BACKGROUND APPLICABLE SAFETY ANALYSES 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 (DBAs).

The OPERABILITY of the accident monitoring instrumentation ensures that there is sufficient information available on selected unit parameters to monitor and to assess unit status and behavior following an accident.

The availability of accident monitoring instrumentation is important so that responses to corrective actions can be observed and the need for, and magnitude of, further actions can be determined. These essential instruments are identified in References 1, 2,[§1 and a[] addressing the recommendations of Regulatory Guide 1.97 (Ref. 3) as required by Supplement 1 to NUREG-0737 (Ref. 4).

The instrument channels required to be OPERABLE by this LCO include two classes of parameters identified during unit specific implementation of Regulatory Guide 1.97 as Type A and Category 1 variables.

These key variables are identified by the unit specific Regulatory Guide 1.97 analyses (Ref. 1, 2,[§1 and a[]). These analyses identify the unit specific Type A and Category 1 variables and provide justification for deviating from the NRC guidance in Reference 3.

The specific instrument Functions listed in Table 3.3.3-1 are discussed in the LCO section.

The PAM instrumentation LCO ensures the OPERABILITY of Regulatory Guide 1.97 Type A variables so that the control room operating staff can:

Perform the diagnosis specified in the emergency operating procedures (these variables are restricted to preplanned actions for the primary success path of DBAs), e.g., loss of coolant accident (LOCA); and Take the specified, pre-planned, manually controlled actions, for which no automatic control is provided, and that are required for safety systems to accomplish their safety function.

Cook Nuclear Plant Unit 2 B 3.3.3-1 Revision No. 16

BASES LCO (continued)

PAM Instrumentation B 3.3.3 One exception to the two channel requirement is Containment Isolation Valve (CIV) Position. In this case, the important information is the status of the containment penetrations. The LCO requires one position indicator for each active CIV. This is sufficient to redundantly verify the isolation status of each isolable penetration either via indicated status of the active valve and prior knowledge of a passive valve, or via system boundary status. If a normally active CIV is known to be closed and deactivated, position indication is not needed to determine status. Therefore, the position indication for valves in this state is not required to be OPERABLE.

Type A and Category 1 variables meet Regulatory Guide 1.97 Category 1 (Ref. 3) design and qualification requirements for seismic and environmental qualification, single failure criterion, utilization of emergency standby power, immediately accessible display, continuous readout, and recording of ~lay, except for approved deviations, as described in References 1 ~

and-0.

Listed below are discussions of the specified instrument Functions listed in Table 3.3.3-1. For all applicable Functions, the recorder or indicator may be used as the qualified instrument.

1.

Neutron Flux Neutron Flux (NRl-21 and NRl-23) is a Category 1 variable provided to verify reactor shutdown. The range of each of the two neutron flux instruments ( 1 0E-8 to 200% power) covers the full range of flux that may occur post accident.

As stated in Note (a) to Table 3.3.3-1, neutron flux instruments are not required to be environmentally qualified to be considered OPERABLE. This is acceptable because verification of initial reactor shutdown is expected to be completed prior to any potential impact to the neutron flux instrumentation due to an adverse containment environment. Other instruments will be used to monitor for subcriticality in the event of an adverse containment environment.

2.

Steam Generator (SG) Pressure (per SG)

Steam Generator Pressure is a Type A, Category 1 variable provided for determination of required core exit temperature. Three steam generator pressure channels per steam generator are provided (MPP-210, MPP-211, MPP-212, MPP-220, MPP-221, MPP-222, MPP-230, MPP-231, MPP-232, MPP-240, MPP-241, and MPP-242).

Each channel has a range of 0 psig to 1200 psig. However, only two steam generator pressure channels per steam generator are required Cook Nuclear Plant Unit 2 B 3.3.3-3 Revision No. 44

BASES LCO (continued)

PAM Instrumentation B 3.3.3 3, 4. Reactor Coolant System (RCS) Hot and Cold Leg Temperatures (Wide Range)

RCS Hot and Cold Leg Temperatures are Type A, Category 1 variables provided for verification of core cooling and long term surveillance. RCS hot and cold le tern eratures are used to determine RCS subcooling margin and to monitor for subcriticalit in he event of an adverse containment environmen.

The RCS hot leg and RCS cold leg channels each receive input from one resistance temperature detector (RTD). In each of RCS loops 1 and 3, there is one RCS hot leg RTD (NTR-110 with MR-9, and NTR-130 with MR-11) and one RCS cold leg RTD (NTR-210 with MR-9, and NTR-230 with MR-11) that satisfy the guidance of Reference 3. The channels provide indication over a range of 0°F to 700°F.

5.

RCS Pressure (Wide Range)

RCS wide range pressure is a Type A, Category 1 variable provided for verification of core cooling and RCS integrity long term surveillance.

RCS wide range pressure is used as criteria to manually trip the reactor coolant pumps.

In addition, RCS wide range pressure is used for determining RCS subcooling margin.

Two RCS Pressure (Wide Range) channels are provided (NPS-110 and NPS-111, with MR-13), each with a range of O psig to 3000 psig.

6.

Reactor Coolant Inventory Tracking System (Reactor Vessel Level Indication)

Reactor coolant inventory is a Category 1 variable provided for verification and long term surveillance of core cooling.

The Reactor Coolant Inventory Tracking System consists of two channels of instrumentation (NLl-110, NLl-111, NLl-120, NLl-121, NLl-130, and NLl-131 ). Each channel is capable of measuring upper plenum level, narrow range level, and dynamic head (i.e., wide range level). The Reactor Coolant Inventory Tracking System provides a direct measurement of the collapsed liquid level above the fuel alignment plate. The collapsed level represents the amount of liquid mass that is in the reactor vessel above the core. Measurement of Cook Nuclear Plant Unit 2 B 3.3.3-4 Revision No. G

BASES LCO (continued}

13.

Steam Generator Water Level (Wide Range)

PAM Instrumentation B 3.3.3 SG Water Level is a Category 1 variable provided to monitor operation of decay heat removal via the SGs. Four steam generator level (wide range} channels (one per steam generator} are provided (BLl-110, BLl-120, BLl-130, and BLl-140). Each channel is capable of monitoring from 12 inches above the steam generator tube sheet to the separators.

14.

Condensate Storage Tank (CST) Level CST Level is a Category 1 variable provided to ensure water supply for auxiliary feedwater (AFW}. The CST provides the qualified water supply for the AFW System. Inventory is monitored from essentially the top of the CST to the bottom of the CST (95% total volume} by a single channel provided to satisfy the guidance of Reference 3, as described in Reference 1. CST Level is displayed on a control room indicator (CLl-114).

15, 16, 17, 18.

Core Exit Temperature Core Exit Temperature is a Type A, Category 1 variable used to determine whether to manually reduce ECCS flow. This variable is also provided for verification and long term surveillance of core cooling. In addition, core exit tern erature is used for determinin RCS subcooling margin and to monitor for subcriticalit in the even fan adverse containment environmen.

Two OPERABLE channels of Core Exit Temperature, with one core exit thermocouple per channel, are required in each quadrant to provide indication of radial distribution of the coolant temperature rise across representative regions of the core. Two core exit temperature channels per quadrant ensure a single failure will not disable the ability to determine the radial temperature gradient. Each core exit temperature channel (SG-30 and SG-31 for TC 1 through 65) has a range of 200°F to 2300°F.

19.

Secondary Heat Sink Indication (per SG)

Cook Nuclear Plant Unit 2 Secondary Heat Sink Indication is a Type A, Category 1 variable used to determine whether to manually reduce ECCS flow. This variable is also provided to monitor operation of decay heat removal via the SGs.

As stated in Note ~

to Table 3.3.3-1, the requirements for this variable are met by any combination of two instruments per SG, B 3.3.3-7 Revision No. Q

BASES PAM Instrumentation B 3.3.3 SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.3.2 Deleted REFERENCES SR 3.3.3.3 CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter with the necessary range and accuracy. This SR is modified by a Note that excludes neutron detectors. For Function 9, the CHANNEL CALIBRATION shall consist of verifying that the position indication conforms to actual valve position. For Functions 15, 16, 17, and 18, whenever a sensing element is replaced, the next required CHANNEL CALIBRATION of the Core Exit Temperature thermocouple sensors is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing elements.

For Functions 20 (Circuit Breaker Status channels) and 24, the CHANNEL CALIBRATION shall consist of verifying that the position indication conforms to actual circuit breaker position. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

1.

NRC letter, T. G. Colburn (NRG) to M. P. Alexich (Indiana Michigan Power Company), "Emergency Response Capability - Conformance to Regulatory Guide 1.97 Revision 3 for the D. C. Cook Nuclear Plant, Units 1 and 2," dated December 14, 1990.

2.

UFSAR, Table 7.8-1.

3.

Regulatory Guide 1.97, Revision 3, May 1983.

4.

NUREG-0737, Supplement 1, "TMI Action Items."

5.

NRG letter, P.S.Tam (NRG), to M. K. Nazar, (Indiana Michigan Power Company), "Donald C. Cook Nuclear Plant, Units 1 & 2 (DCCNP-1 and DCCNP-2) - Issuance of Amendments Re: Containment Sump Modifications per Generic Letter 2004-02 (TAC Nos. MD5901 and MD5902)," dated October 18, 2007.

6.

Letter from Indiana Michigan Power Company (K. J. Ferneau) to the NRC dated XXXX, XX 2024.

7.

NRC letter, X. X. XXXXXX (NRG), to Q. S. Lies (Indiana Michigan Power Company), XXXX, dated XXXX, XX 2024.

Cook Nuclear Plant Unit 2 B 3.3.3-15 Revision No. eO