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| {{#Wiki_filter:Indiana Michigan Power INDIANA Cook Nuclear Plant MICHIGAN One Cook Place POWER Bridgman, Ml 49106 indianamichiganpower.com An AEP Company BOUNDLESS ENERGY-August 2, 2023 AEP-NRC-2023-34 10 CFR 50.90 Docket Nos.: 50-315 50-316 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Donald C. Cook Nuclear Plant Unit 1 and Unit 2 SUPPLEMENT TO REQUEST FOR APPROVAL OF CHANGE REGARDING NEUTRON FLUX INSTRUMENTATION | | {{#Wiki_filter:}} |
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| ==Reference:==
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| : 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 Accession (ADAMS) No. ML23026A284.
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| This letter is provided by Indiana Michigan Power Company (l&M), licensee for Donald C. Cook Nuclear Plant (CNP) Unit 1 and Unit 2, as a supplement to Reference 1 regarding a request to reclassify the wide range neutron flux instrumentation at CNP Unit 1 and Unit 2 as Category 3 instrumentation, as defined by Regulatory Guide 1.97, Revision 3. This request would modify CNP Unit 1 and Unit 2 Technical Specifications Table 3.3.3-1, Post Accident Monitoring Instrumentation, to remove Function 1, Neutron Flux, from the list of required post-accident monitoring instrumentation.
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| During a conference call between l&M and U.S. Nuclear Regulatory Commission (NRC) staff, held on June 14, 2023, it was determined that supplemental information would be provided to support the NRC staffs review of the amendment request. This supplement provides information on the design and operation of existing instrumentation, as well as how the proposed changes would continue to provide control room operators with the information required to safely operate the plant.
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| Enclosure 1 to this letter provides an affirmation statement. l&M is providing Enclosure 2 to this letter as a supplement to the letter provided by l&M as Reference 1.
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| U.S Nuclear Regulatory Commission AEP-NRC-2023-34 Page 2 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.
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| Sincerely,
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| .2 Senior Vice President and Chief Nuclear Officer BMC/kmh
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| ==Enclosures:==
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| : 1. Affirmation
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| : 2. Supplement to Request for Approval of Change Regarding Neutron Flux Instrumentation c: EGLE -- RMD/RPS J. B. Giessner - NRC Region Ill M. G. Menze -- AEP Ft. Wayne, w/o enclosures NRC Resident Inspector N. Quilco -- MPSC R. M. Sistevaris -AEP Ft. Wayne, w/o enclosures S. P. Wall-NRG Washington, D.C.
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| A. J. Williamson -- AEP Ft. Wayne, w/o enclosures
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| Enclosure 1 to AEP-NRC-2023-34 AFFIRMATION I, Q. Shane Lies, being duly sworn, state that I am the Senior Vice President and Chief Nuclear Officer 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.
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| Indiana Michigan Power Company 2.%<5 Senior Vice President and Chief Nuclear Officer SWORN TO AND SUBSCRIBED BEFORE ME mis_[)_oA or hugust 2o23 re oiry My Commission Expires ( ]1/205
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| Enclosure 2 to AEP-NRC-2023-34 Supplement to Request for Approval of Change Regarding Neutron Flux Instrumentation By {{letter dated|date=January 26, 2023|text=letter dated January 26, 2023}} (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 reclassify the wide range (WR) neutron flux instrumentation at CNP Unit 1 and Unit 2 as Category 3 instrumentation, as defined by Regulatory Guide (RG) 1.97, Revision 3 (Reference 2). This request would modify CNP Unit 1 and Unit 2 Technical Specifications Table 3.3.3-1, Post Accident Monitoring Instrumentation, to remove Function 1, Neutron Flux, from the list of required post-accident monitoring instrumentation.
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| During a conference call between l&M and U.S. Nuclear Regulatory Commission (NRC) staff, held on June 14, 2023, it was determined that supplemental information would be provided to support the NRC staffs review of the amendment request. This supplement provides information on the design and operation of existing instrumentation, as well as how the proposed changes would continue to provide control room operators with the information required to safely operate the plant.
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| ===Background===
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| Following the partial core meltdown at Three Mile Island Unit 2 (TMI), in March of 1979, a number of new requirements were established for all operating nuclear power plants in an effort to minimize the possibility of a similar event occurring at another nuclear power plant. One action that was required following the TMI incident, as discussed in NUREG-0737 (Reference 3) and its supplement (Reference 4), was the development of new, symptom-based Emergency Operating Procedures (EOPs). Each reactor vendor developed generic Emergency Response Guidelines (ERGs), written to a defined set of typical plant configurations that best approximated the existing fleet, and individual nuclear plants developed site-specific EOPs based on their associated ERGs.
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| CNP Unit 1 and Unit 2 are both Westinghouse Pressurized Water Reactors. As such, the CNP Unit 1 and Unit 2 EOPs are based on ERGs developed by Westinghouse Electric Company, LLC. In June of 1983, the NRC issued Generic Letter 83-22 (Reference 5) to document the safety evaluation performed for the Westinghouse ERGs, stating that the guidelines are acceptable for implementation and will provide improved guidance for EOP development.
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| In December of 1980, the NRC issued Revision 2 of RG 1.97, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident (Reference 6). This was followed by Revision 3 of RG 1.97 in May of 1983 (Reference 2). 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.
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| In June of 1987, l&M submitted a letter indicating a decision to install environmentally qualified WR neutron flux instrumentation at CNP Unit 1 and Unit 2 during the refueling outages scheduled to occur
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| Enclosure 2 to AEP-NRC-2023-34 Page 2 in 1989, as part of compliance with RG 1.97 recommendations (Reference 7). The WR neutron flux instrumentation installed at CNP Unit 1 and Unit 2 (referred to as Gamma-Metrics or WR log power) is comprised of fission chambers located exterior to the reactor vessel, in a location similar to that of the existing Westinghouse nuclear instrumentation (source range, intermediate range, and power range detectors). The Gamma-Metrics detectors rely on thermal neutrons leaking from the core in order to provide an indication of in-core activity.
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| The CNP Unit 1 and Unit 2 EOPs were revised to include the use of the WR neutron flux instrumentation, but the Westinghouse ERGs were not revised as such, and reactor condition within the ERGs continued to be assessed using the originally-installed source range, intermediate range, and power range nuclear instrumentation. l&M documented this deviation from the Westinghouse ERGs in the plant specific background document for the associated EOPs.
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| Current Operation Within the CNP Unit 1 and Unit 2 EOPs, the Critical Safety Function Status Tree (CSFST) for Subcriticality, OHP-4023-F-0.1 (F-0.1 ), directs control room operators to use WR neutron flux instruments to monitor for unexpected additions of reactivity after reactor shutdown has been achieved.
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| This status tree directs operators to enter function restoration procedure OHP-4023-FR-S.1 (FR-S.1 ),
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| Response to Nuclear Power Generation/ATWS, in the event that Gamma-Metrics indicate WR log power is 5 percent (%) or greater, or if WR log power is greater than 10% and WR start up rate is positive. FR-S.1 provides actions to add negative reactivity, including emergency boration, to a core that is observed to be critical when expected to be shutdown. Operators are directed to continue use of FR-S.1 until WR log power is less than 5% and WR start up rate is negative, and to continue boration to maintain adequate shutdown margin during subsequent recovery actions. It should be noted that the plant specific background document for F-0.1 states that the 5% WR log power value "was chosen because it is clearly readable on the wide range log power meter."
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| In the event that the CSFST for Subcriticality denotes a yellow condition (WR log power greater than 10% and start up rate is between -0.2 decades per minute (DPM) and O DPM, or WR log power less o-than 1 5% with a positive WR start up rate), operators are directed to enter function restoration procedure OHP-4023-FR-S.2 (FR S.2), Response to Loss of Core Shutdown. This procedure directs operators to emergency borate until WR log power is less than 10% and WR start up rate is zero or negative, and to continue boration to maintain adequate shutdown margin during subsequent recovery actions.
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| Emergency boration at CNP Unit 1 and Unit 2 involves starting boric acid transfer pumps and placing them in fast speed, then aligning the boric acid flowpath to the charging pump suction for injection into the reactor coolant system (RCS). Both FR-S.1 and FR-S.2 direct operators to verify boration flow of at least 44 gallons per minute (gpm) using instrument QFl-410. If this flowpath cannot be verified, operators are directed to use an alternate flowpath, via the boric acid blender, and to verify boration flow of at least 36 gpm using instrument QFC-411. FR-S.1 also allows for operators to borate by aligning at least one charging pump to the refueling water storage tank, maximizing charging flow, and verifying letdown flow is established. Each of the instruments discussed above is powered from vital instrument power supplies and will be available during all postulated accident conditions, including the boric acid transfer pump running lights.
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| Enclosure 2 to AEP-NRC-2023-34 Page 3 If the emergency core cooling system (ECCS) is in service, via a safety injection signal, then the flow rates of each set of injection pumps are available for the operator to monitor for boric acid injection as follows:
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| * Boron injection flowpath - IFl-51, IFl-52, IFl-53, IFl-54 (high head, per loop injection)
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| * Safety injection flowpath - IF1-260, IF1-266 (medium head)
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| * Residual heat removal flowpath- lFl-310, IFl-311, IFl-320, IFl-321 (low head)
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| Each of these flow rates is monitored within the EOPs following a safety injection actuation.
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| Additionally, monitor lights are provided at the top of the control panels to indicate valve positions for the various signals to ensure that ECCS flowpaths are properly aligned.
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| Operators are trained to establish stable conditions for long term heat removal in the event of an accident scenario. Once a stable balance of temperatures and flows has been established, any changes that occur that were not initiated or expected by control room operators will be promptly investigated and corrected, per established training and the guidance of procedures in progress.
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| Evaluation of the Requested Change During the evaluation of the current conditions at CNP Unit 2, with one train of WR log power not able to meet its environmental qualification requirements, it was observed that when performing FR-S.1, operators rely on Gamma-Metrics as the sole input in the decision to transition back to optimal recovery procedures. In the event that both Gamma-Metrics are rendered inoperable, this procedural requirement could result in operators not exiting FR-S.1 even if all other relevant parameters indicate that the reactor is shut down. As such, training has been provided to operators, and a procedure change is being processed, outside of the requested change, to allow control room operators to determine reactor condition using an aggregate evaluation of several instruments, including core exit thermocouple temperatures, hot and cold leg temperatures, any indicating nuclear instruments, and reactor boron concentration. If an aggregate indication review determines that the reactor is subcritical, control room operators will be directed to continue boration to maintain adequate shutdown margin and return to the procedure and step in effect.
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| As discussed in the original license amendment request, during an accident that involves normal containment conditions, neutron flux instrumentation is expected to be available for use, regardless of whether it meets environmental qualification criteria. Thus, reclassifying WR neutron flux instruments as Category 3 instrumentation would have no effect on operator response to any accident involving normal containment conditions.
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| In an accident scenario that results in an adverse containment, WR neutron flux instrumentation, if not environmentally qualified, cannot be relied upon to function for the duration of the event. In an accident scenario where WR neutron flux instruments are rendered inoperable, operators would be directed to enter FR-S.1 since at that point WR log power less than 5% cannot be confirmed. Control room operators, as directed by FR-S.1, would initiate actions to add negative reactivity to the core, including emergency boration, and would remain in FR-S.1 until it can be determined, using an aggregate indication review, that the reactor is subcritical.
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| Enclosure 2 to AEP-NRC-2023-34 Page 4 Main Steam Line Break During operating conditions when the RCS is intact with subcooling, the time rate of change of neutron flux ( startup rate) can be used as a single variable that would indicate criticality or a trend toward criticality. The only credible accident condition that would result in both adverse containment and an intact RCS is a main steam line break (MSLB) inside containment. In the event of a MSLB inside containment where all nuclear instrumentation is rendered inoperable, the Critical Safety Function of Subcriticality can still be assessed using existing environmentally qualified instrumentation.
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| The fundamental characteristic of a MSLB is a rapid cooldown and depressurization of the intact RCS due to the uncontrolled heat removal via the high blowdown steam flow out of the break. The steam generator blowdown causes a rapid pressure decrease in the faulted steam generator, which initiates a reactor trip signal and safety injection actuation. The rapid RCS cooldown causes a positive reactivity insertion due to the negative moderator temperature coefficient and causes a return to power. Since it is a design basis condition to expect a return to power following a MSLB, automatic protection equipment is provided in the form of emergency boration via the safety injection system. Thus, no operator action is required to mitigate this expected initial reactivity transient.
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| At some point during the MSLB event, the adverse containment environment may cause the Gamma-Metrics to become inoperable. With no WR log power indications available to control room operators, operator training and the F-0.1 procedure would direct operators to enter FR-S.1. Revisions to the FR-S.1 procedure would instruct operators to use an aggregate indication, as discussed above, to determine whether the reactor is subcritical, and would direct operators to emergency borate in the event of a return to criticality. Once operators successfully exit the FR-S.1 procedure, operator training and changes to the background document for F-0.1 would allow control room operators to rely on an aggregate indication of plant instruments, including core exit thermocouples (CET), RCS hot and cold leg temperatures, and reactor boron concentration, as they continue to evaluate for subcriticality.
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| In the unlikely and unpredicted event that a return to power were to occur from an unknown boron dilution that may be in progress after the MSLB, the core average temperature would increase due to the increase in core heat flux caused by the generation of nuclear power. With the CETs located inside the reactor vessel and mounted directly above the active fuel region within the coolant flow path, any rise in heat output from the fuel assemblies will be seen by the CETs within seconds of the change.
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| Control room operators will have established a heat balance with known auxiliary feedwater injection rates at a stable supply temperature, along with a heat removal steam flow that they control and monitor.
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| If the containment conditions are such that reactor coolant pump (RCP) operation is still allowed, then the CETs will see the change in temperature output from the fuel almost immediately. If natural circulation conditions have been established because the RCPs had to be secured, the coolant flow will be lower, but still high enough to provide a timely indication at the CETs. The operators would then enter FR-S.1 and initiate emergency boration of the RCS until it can be determined, using the CETs, RCS hot and cold leg instruments, and boron concentration, that the reactor is subcritical.
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| Loss of Coolant Accident Besides a MSLB inside containment, all other conditions that result in in adverse containment involve an RCS that is not intact. The fundamental characteristic of a loss of coolant accident (LOCA) is a depressurization of the RCS and a pressurization of containment. A safety injection actuation signal is generated when the appropriate low pressurizer pressure setpoint is reached, limiting the consequences of the LOCA by causing a rapid power reduction and providing heat transfer from the
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| Enclosure 2 to AEP-NRC-2023-34 Page 5 core. The reactor trip and subsequent injection of borated water maintain subcriticality throughout the scenario. If reactor trip cannot be verified, even after manually tripping the reactor, the EOPs direct the operator to enter FR-S.1 and commence emergency boration of the RCS.
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| During the recirculation phase of a LOCA event, dilution of the borated inventory in containment can occur from sources such as a containment spray heat exchanger tube leak. Such an event would develop slowly due to the large volume of borated water transported into containment during the injection phase of the accident. This allows sufficient time to observe the dilution with ongoing chemistry sampling and temperature feedback.
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| The blowdown of the RCS associated with a LOCA can cause significant core voiding, core uncovery, and reactor vessel downcomer uncovery. The core voiding and core uncovery caused by RCS blowdown will heavily influence neutron flux instrumentation response. As the ECCS and two-phase RCS mixture is pumped through the downcomer and core, three effects are manifest: (1) less water (moderator) in the core decreases the intrinsic neutron source reading due to an increase in fast neutrons vs thermal neutrons; (2) decreased fluid density in the downcomer permits more neutrons to leak out to the excore detectors; (3) increased leakage from the core and decreased moderator density reduce neutron multiplication. Although fewer thermal neutrons remain in the core to help sustain the fission process, many more are able to escape to the neutron detector for measurement. Therefore, in a voided or uncovered core the neutron flux readings could be misleading and imply a high neutron flux level when, in fact, the core is effectively shut down. Under these circumstances, core exit temperature would more appropriately monitor the status of the core since there would be no heat generated by nuclear power in a shut down core. Additionally, it is important to recognize that voiding different regions of the core will have a varying effect on excore detector readings.
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| As mentioned in the original license amendment request (Reference 1), the above phenomena of erroneous flux indication was observed on the installed Westinghouse nuclear instrumentation during the 1979 TMI incident (Reference 8), and has further been demonstrated in NRC financed experiments at the Pennsylvania State University reactor and LOFT facility (Reference 9).
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| Thus, throughout a LOCA event, WR log power alone is typically inadequate for early detection of criticality due to the unreliable indication caused by a voided RCS. Control room operators are aware of this limitation and are trained to monitor multiple indications and to assess for aggregate for indications of criticality. These include CET trends, RCS hot and cold leg temperature indications, and boron sample results. ECCS flows and temperatures will be monitored by control room operators, and, due to the position of the CETs, a change in core exit temperatures would be observed quickly. Control room staff would conclude that a rising temperature trend that is not accompanied by a reduction in ECCS flow or ECCS cooling flow would be considered an indicator of a reactivity concern requiring operator action during all accident conditions. In the event that core voiding has occurred, the increase in heat output would also be identified by containment pressure indications due to the increased steam flow from the core.
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| While RCS hot and cold leg temperature readings will be somewhat delayed compared to the CETs, the use of trending will still allow for sufficiently early detection of a return to power event. Furthermore, this value is relatively insensitive to voiding in comparison to the excore instrumentation, as discussed above. Therefore, the use of CETs and RCS hot and cold leg temperature instruments provides a method of detection that is just as reliable as excore instrumentation alone in the event of a LOCA.
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| to AEP-NRC-2023-34 Page6 Conclusion In the case of an adverse containment, whether a MSLB or a LOCA, the proposed method of using aggregate indication, including CETs, RCS hot and cold leg temperatures, and reactor boron concentration provides control room operators sufficiently reliable and meaningful information to determine if a return to power is underway.
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| ==References:==
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| : 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 Number ML23026A284).
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| : 2. 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 Number ML003740282).
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| : 3. NUREG-0737, "Clarification of TMI Action Plan Requirements," Revision 0, November 1980 (ADAMS Accession Number ML051400209).
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| : 4. NUREG-0737, Supplement No. 1, "Clarification of TMI Action Plan Requirements,"
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| Revision 0, January 1983 (ADAMS Accession Number ML102560009).
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| : 5. Generic Letter 83-22, "Safety Evaluation of "Emergency Response Guidelines,"" dated June 3, 1983.
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| : 6. 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 Number ML060750525).
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| : 7. Letter from M. P. Alexich, l&M, to Dr. T. E. Murley, NRC, "Additional Information on and Requests for Deviations from Regulatory Guide 1.97, Rev. 3 Recommendations," dated June 29, 1987 (ADAMS Accession Number ML173348114).
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| : 8. NSAC-1, "Analysis of Three Mile Island - Unit 2 Accident," Electric Power Research Institute (Nuclear Safety Analysis Center), July 1979.
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| : 9. The Response of Ex-Core Neutron Detectors to Large- and Small-Break Loss-of-Coolant Accidents in Pressurized Water Reactors, E. Wilson Okyere, Anthony J. Baratta, and William A. Jester, Nuclear Technology, Volume 96, December 1991.}}
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