Text

License Amendment Request to Extend the Completion Time for Sodium Hydroxide System
	ML25140A128
Person / Time
Site:	Ginna
Issue date:	05/20/2025
From:	Knowles J; Constellation Energy Generation
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	Download: ML25140A128 (1)
	v • d • e

200 Energy Way Kennett Square, PA 19348 www.constellation.com Constellation May 20, 2025 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 R.E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 NRC Docket No. 50-244 10 CFR 50.90

Subject:

License Amendment Request to Extend the Completion Time for Sodium Hydroxide System In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG) requests an amendment to Renewed Facility Operating License No. DPR-18 for R.E. Ginna Nuclear Power Plant (Ginna).

The proposed change extends the Completion Time for Technical Specification (TS) 3.6.6 "Containment Spray (CS), Containment Recirculation Fan Cooler (CRFC), and NaOH Systems,"

Required Action B Sodium Hydroxide (NaOH) system inoperable. Specifically, the proposed change extends the Required Action B Completion Time from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 14 days. This TS change will allow for maintenance to be performed online and eliminate the potential need for Notice of Enforcement Discretion (NOED) in the future. proposes several administrative changes; adding notes for clarification, removing notes that no longer apply, removal of words that no longer apply following previous amendments, and formatting and editorial changes.

A note will be added to Surveillance Requirement (SR) 3.4.14.2 consistent with a regulatory commitment. The note will clarify that Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) check valve pairs are allowed to be tested as a single valve.

The note included in TS 3.4.7, 3.4.8, 3.9.4 and 3.9.5 that allowed an alternative means of Residual Heat Removal (RHR) that was valid between April 3, 2020 and June 30, 2020 is no longer valid and will be removed.

The Spent Fuel Pool (SFP) Charcoal Adsorber System will be removed from the first paragraph of TS 5.5.10. The SFP Charcoal Adsorber System was removed from the Ventilation Filter Testing Program (VFTP) in a previous amendment however SFP Charcoal Adsorber was not removed from the first paragraph of TS 5.5.10.

The Surveillance Requirements table containing SR 3.7.10.1 and 3.7.10.2 will have a vertical line added between the surveillance column and the frequency column. A horizontal line will be removed between the note and the Surveillance in SR 3.4.13.2. Typographical errors will be corrected in TS 3.1.6, SR 3.2.1.2, and TS 3.2.3.

The proposed changes have been reviewed by the Ginna Plant Operations Review Committee in accordance with the requirements of the CEG Quality Assurance Program.

This amendment request contains no new regulatory commitments.

Page 2 CEG requests approval of the proposed amendment by May 20, 2026. Once approved, the amendment shall be implemented within 60 days.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b), CEG is notifying the State of New York of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

Should you have any questions concerning this letter, please contact Michael Henry at 267-533-5382.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 20th day of May 2025.

Respectfully,

Knowles, Justin W Justin Knowles Digitally signed by Knowles, Justin W Date: 2025.05.20 11 :00:15 -04'00' Senior Manager - Licensing Constellation Energy Generation, LLC Attachments:

1. Evaluation of Proposed Change

2. Markup of Technical Specifications Pages for Technical Specification 3.6.6

3. Markup of Technical Specifications Bases Pages for Technical Specification 3.6.6

4. Evaluation of Proposed Editorial Changes

5. Markup of Technical Specifications Pages for Editorial Changes

6. Markup of Technical Specifications Bases Pages for Editorial Changes cc:

NRC Regional Administrator, Region I NRC Senior Resident Inspector, Ginna NRC Project Manager, Ginna A. L. Peterson, NYSERDA A. Kauk, NYSPSC

ATTACHMENT 1 Evaluation of Proposed Changes

Subject:

License Amendment Request to Extend the Completion Time for Sodium Hydroxide System 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 System Design and Operation 2.2 Current Technical Specification Requirements 2.3 Reason for the Proposed Change

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 No Significant Hazards Consideration 4.3 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG) requests an amendment to Appendix A, Technical Specifications (TS) of Renewed Facility Operating License DPR-18 for R.E. Ginna Nuclear Power Plant (Ginna).

The proposed change extends the Completion Time for TS 3.6.6 "Containment Spray (CS),

Containment Recirculation Fan Cooler (CRFC), and NaOH Systems" Condition B "NaOH system inoperable" from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 14 days.

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation The Sodium Hydroxide (NaOH) system at Ginna is injected into the Containment Spray (CS) flowpath via a liquid eductor during the injection phase of an accident. The NaOH system primarily consists of one tank, two air operated valves (AOVs), two check valves, and two eductors. The only primary components that are active include the two AOVs in parallel, all remaining primary components are considered passive. The AOVs are normally closed and will open upon receiving the proper safety signal. Additionally, the AOVs fail open on loss of air. The eductors ensure that the pH of the spray mixture is a caustic solution. The NaOH added in the spray ensures an alkaline pH for the solution recirculated in the containment sump. The alkaline pH of the containment sump water minimizes the evolution of iodine and minimizes the occurrence of chloride and caustic stress corrosion on mechanical systems and components exposed to the fluid.

2.2 Current Technical Specification Requirements The current requirement of TS 3.6.6 "Containment Spray (CS), Containment Recirculation Fan Cooler (CRFC), and NaOH Systems" Condition B "NaOH system inoperable" is to restore the NaOH system to operable within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Failure to restore the NaOH system to operable within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months requires Condition C "Required Action and associated Completion Time of Condition A or B not met. Requires a unit shutdown, be in Mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and be in Mode 5 in 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months .

The operability of the NaOH system aids in scavenging fission products from the containment atmosphere during a Design Basis Loss of Coolant Accident (DBA) and ensures an alkaline pH for the solution recirculated in the containment sump.

2.3 Reason for the Proposed Change The proposed NaOH System Completion Time extension would provide enhanced safety benefits by supporting the ability to complete on-line corrective or planned maintenance on the NaOH System. This proposed amendment will reduce the potential for unplanned unit shutdowns. It will increase the time to perform troubleshooting, repair and testing following NaOH system issues, which will enhance the safety and reliability of the equipment. This amendment will also allow time to perform routine maintenance on the NaOH System online.

The proposed completion time extension does not significantly increase the potential for the NaOH system being inoperable when required to actuate with the containment spray system

during a Large Break Loss of Coolant Accident (LOCA) event. The probability of a Large Break LOCA defined in the Ginna Initiating Events PRA Notebook is 6.33E-6/year (approximately one occurrence in 158,000 years). Additionally, this proposal would minimize the potential need for NOED.

3.0 TECHNICAL EVALUATION

Design Basis Summary The CRFC System, CS System and NaOH System are Engineered Safety Feature (ESF) systems. They are designed to remove sufficient heat from the containment atmosphere following an accident condition to maintain the containment pressure below design limits. The containment spray system in conjunction with the NaOH system is also capable of reducing the iodine and particulate fission product inventories in the containment atmosphere such that the offsite radiation exposure resulting from a LOCA is within the guidelines established by 10 CFR 50.67. In addition to depressurization and removal of fission products, the NaOH system is also responsible for increasing the emergency core cooling system (ECCS) solution pH. This reduces the likelihood of stress corrosion cracking of stainless steel components. The addition of NaOH ensures that the containment sump liquid pH range is greater than the minimum required value of 7.0 as specified in Branch Technical Position MTEB 6-1, and less than 10.5 as specified in Standard Review Plan 6.5.2.

The CS System operates independently of the NaOH system, the NaOH system inoperability has no impact on the operability of the CS system. The CS system will operate to reduce and maintain the design pressure within containment regardless of the NaOH system.

The CRFC and filtration system provides a dynamic heat sink to cool the containment atmosphere and filtration of the containment atmosphere to remove airborne particulate and iodine fission products that are the source for potential public exposure. The system utilizes the normal containment ventilation and cooling equipment in addition to the charcoal filters. In the event of an accident, the flow of two CRFCs would be directed through an alternative bypass line to the post-accident charcoal filters before being discharged onto the operating floor area of containment, which removes airborne particulates and iodine from the containment atmosphere.

These redundant features function independently of the NaOH system and act to minimize containment pressure and reduce the airborne particulate and iodine in the containment atmosphere.

Probability Insights A review of the Ginna PRA model determined that the NaOH System is not explicitly modeled in the Ginna Full Power Internal Events (FPIE) or Fire PRA models. Per the PRA Containment Spray System Notebook, G1-PRA-005.006, "The blending of sodium hydroxide (NaOH) from the spray additive tank with the Containment Spray flow to the spray headers is not included in the system model. This function is needed to reduce the elemental iodine in the containment atmosphere such that the offsite radiation exposure resulting from a LOCA is within the guidelines established by 10 CFR 50.67. This function is not required for containment pressure control which is of interest for the Level 1 and 2 PRA." The NaOH System is also not credited in the PRA Large Early Release Analysis Notebook, G1-PRA-015. The NaOH System does not mitigate core damage or a large early release.

The NaOH System is required to actuate and be an additive to the Containment Spray in the event of a LOCA. The probability of a Large Break LOCA, from the Ginna Initiating Events PRA Notebook, is 6.33E-6/year (approximately one occurrence in 158,000 years).

Since the NaOH system is not modeled in the Ginna FPIE or Fire PRA, it would not impact the Average Annual Core Damage Frequency (CDF) or Large Early Release Frequency (LERF) for Ginna. Therefore, PRA cannot calculate any change in CDF or LERF based on the duration of unavailability of the NaOH system. However, the likelihood of experiencing a Design Basis Large Break LOCA that would require the actuation of Containment Spray and the NaOH System in the Ginna Initiating Events PRA Notebook is 6.33E-6/year (approximately one occurrence in 158,000 years). This probability of occurrence provides reasonable assurance that the 14 day completion time is appropriate.

Impact to Alternative Source Term (ASTI Loss of Coolant Accident (LOCA) Analysis Additional defense in depth was assessed by considering the impact on the AST LOCA Analysis. The proposed change does not impact the design basis or accident analysis. The proposed change to the NaOH completion time will be evaluated by reviewing the probability and consequence of the limiting design basis accident with an extension from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 14 days. CEG performed a technical evaluation of the impact to the AST analysis of record to assess the dose impact if an accident were to occur with NaOH system unavailable.

The Ginna AST Loss of Coolant Accident (LOCA) analysis models two distinct release paths following a LOCA: Containment leakage, and ECCS leakage. The first release path is containment leakage which assumes the air in containment will leak directly to the environment.

This release path takes credit for a portion of the Containment Volume being sprayed by Containment Spray with NaOH additive included for elemental iodine removal. The second release path is the ECCS leakage case. The ECCS leakage in this release path is outside containment and leaks directly to the environment. This case is not impacted by changes to Containment Spray. The limiting results involve the summation of both dose cases and the inclusion of shine calculation to the Control Room dose results. The shine dose results would not be expected to be impacted by the change in NaOH Spray availability.

For the Containment leakage case, the containment is broken up into two distinct volumes. The "Unsprayed" compartment accounts for the portion of Containment that is not impacted by Containment Spray. In this compartment, Aerosol modeling is accounted for by forced recirculation flow by Fan Coolers with HEPA and Carbon Filters. The modeling of Containment Spray does not impact that compartment.

The "Sprayed" compartment accounts for the remaining portion of Containment that is directly impacted by Containment Spray. The overall containment spray system is used to reduce the concentrations of particulate and elemental iodine in the Containment atmosphere. The liquid spray traps particulates while the NaOH binds to elemental iodine. Removal of the NaOH additive system would eliminate the spray's ability to remove elemental iodine. It would not impact the availability of the Containment Spray system, nor would it impact the flowrates and other capabilities of containment spray. Therefore, to account for the impact to the AST LOCA Dose results, the Containment Leakage release path was evaluated without crediting the elemental iodine removal capabilities modeled in the "Sprayed" compartment.

The dose calculations are performed using the RADTRAD computer code version 3.03 in the same manner as the Ginna Analysis of Record (AOR). Results of this analysis show that sufficient margin remains to the Regulatory Guide (RG) 1.183 Revision O acceptance criteria for the duration of the event if it were to occur when NaOH additive system is unavailable. The

EAB LPZ Control Room Containment 2.9822 0.98074 2.3108 ECCS 0.1239 0.1973 1.638 Shine 0.36 AOR 3.11 1.18 4.31 Limit 25 25 5

EAB LPZ Control Room Containment 3.1714 1.2207 2.9217 ECCS 0.1239 0.1973 1.638 Shine 0.36 AOR 3.30 1.42 4.92 Limit 25 25 5

results of the official AOR are shown in Table 1. The AST LOCA evaluation results with NaOH system unavailable for Iodine removal are in Table 2.

Table 1: Ginna AST LOCA AOR Results Table 2: Ginna AST LOCA Evaluation with NaOH System Unavailable Summary and Conclusion The engineered safety features protection systems at Ginna have sufficient redundancy of component and power sources such that under the conditions of a design basis LOCA, the system can, even in the event of a single failure, maintain emergency core cooling, maintain the integrity of the containment, and perform other safeguards functions to ensure that post accident exposures are maintained below the guidelines of 10 CFR 50.67. The systems that will function redundant to the NaOH system are the CS system, which acts to lower containment pressure, and the CRFCs, which act to reduce pressure and remove airborne particulates and iodine from the containment atmosphere. Additional margin is shown in the evaluation of the AST LOCA Evaluation without crediting NaOH. The AST LOCA Evaluation demonstrates that sufficient margin remains to the RG 1.183 Revision O acceptance criteria for the duration of the event if it were to occur when NaOH additive system is unavailable.

NaOH is not modeled in the PRA model therefore it does not impact CDF or LERF. The frequency of the large break LOCA from the Ginna Initiating Events PRA Notebook is 6.33E-6/year (approximately one occurrence in 158,000 years), and this probability of occurrence provides reasonable assurance that the 14 day completion time is appropriate.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.36(c) provides that TS will include Limiting Conditions for Operation (LCOs), which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO of a nuclear reactor is not met, the licensee will shut down the reactor or follow any remedial action permitted by the TS until the condition can be met. The proposed change involves extending the Completion Time for TS 3.6.6, Required Action B, from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 14 days. The LCO itself remains unchanged, as do the required remedial actions or shut down requirements in accordance with 10 CFR 50.36. In addition, 10 CFR 50.36 requires that a licensee's TS be derived from the analyses and evaluation included in the safety analysis report. The proposed change does not affect compliance with the intent of 10 CFR 50.36.

10 CFR 50.67, "Accident source term," allows certain licensees to voluntarily revise the accident source term used in design basis radiological consequence analyses. The AST methodology described in 10 CFR 50.67 has been adopted at Ginna using the methodology of RG 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," Revision 0. The NaOH system is designed to limit the Total Effective Dose Equivalent (TEDE) within the guidelines of 10 CFR 50.67 by reducing the dose contribution from Iodine. The proposed change provides an extension to the Completion Time for TS 3.6.6, Required Action B, and does not involve a change to the OBA LOCA analysis for Ginna. In addition, a radiological evaluation has concluded that the regulatory limits for offsite and control room dose continue to be met with the NaOH system unavailable. The proposed change does not affect compliance with the intent of 10 CFR 50.67.

4.2 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG) requests an amendment to Renewed Facility Operating License No. DPR-18 for R.E. Ginna Nuclear Power Plant (Ginna).

The proposed change extends the Completion Time for Technical Specification (TS) 3.6.6 "Containment Spray (CS), Containment Recirculation Fan Cooler (CRFC), and NaOH Systems,"

Required Action B Sodium Hydroxide (NaOH) system inoperable. Specifically, the proposed change extends the Required Action B Completion Time from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 14 days. This TS change will allow for maintenance to be performed online and avoid unnecessary unit shutdowns and eliminate the potential need for Notice of Enforcement Discretion (NOED) in the future.

R.E. Ginna Nuclear Power Plant has evaluated whether a significant hazards consideration is involved with the proposed amendment 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 change to the NaOH completion time does not involve a significant increase in the probability or consequence of an accident previously evaluated because there is not a significant change to the design or operation of the NaOH system. NaOH system inoperability is not an initiator to any analyzed events. Additionally, TS actions and the associated completion time are not initiators of previously evaluated accidents. Extending the completion time for an

inoperable NaOH system would not have a significant impact on the occurrence of an accident previously evaluated. The proposed amendment will not result in modifications to plant activities associated with NaOH system maintenance. It will provide operational flexibility, allowing additional time to perform system troubleshooting, corrective maintenance, and post-maintenance testing online.

Therefore, the proposed changes do 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 amendment does not involve physical alteration of the NaOH System or any other system at Ginna. No new equipment is being introduced, and installed equipment is not being operated in a new or different manner. There is no change being made to the parameters within which Ginna is operated. There are no setpoints at which protective or mitigating actions are initiated that are affected by this proposed action. The change does not alter assumptions made in the safety analysis. No alteration is proposed to the procedures that ensure Ginna remains within analyzed limits, and no change is being made to procedures relied upon to respond to an off-normal event. As such, no new failure modes are being introduced.

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 Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The proposed change, which would increase the completion time from 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 14 days for NaOH System inoperability, does not exceed or alter a setpoint, design basis or safety limit.

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

Based on the above, CEG concludes that the proposed amendment does not involve a 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.

4.3 Conclusions 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 Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Markup of Technical Specifications Pages for Technical Specification 3.6.6 R.E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 REVISED TECHNICAL SPECIFICATIONS PAGE 3.6.6-1

CS, CRFC, and NaOH Systems 3.6.6 3.6 CONTAINMENT SYSTEMS 3.6.6 Containment Spray (CS), Containment Recirculation Fan Cooler (CRFC), and NaOH Systems LCO 3.6.6 Two CS trains, four CRFC units, and the NaOH system shall be OPERABLE.

-NOTE-In MODE 4, both CS pumps may be in pull-stop for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for the performance of interlock and valve testing of motor operated valves (MOVs) 857 A, 857B, and 857C. Power may also be restored to MOVs 896A and 896B, and the valves placed in the closed position, for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for the purpose of each test.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One CS train inoperable.

A.1 Restore CS train to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status.

OR

NOTE--------

Not applicable if there is a loss of function.

In accordance with the Risk Informed Completion Time Program B.

NaOH system inoperable.

8.1 Restore NaOH System to 172 hours0.00199 days 0.0478 hours 2.843915e-4 weeks 6.5446e-5 months !(

114 d OPERABLE status.

ays C.

Required Action and C.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion AND Time of Condition A or B not met.

C.2 Be in MODE 5.

84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months R.E. Ginna Nuclear Power Plant 3.6.6-1 Amendment 458--

Markup of Technical Specification Bases Pages for Technical Specification 3.6.6 R.E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 REVISED TECHNICAL SPECIFICATION BASES PAGE B 3.6.6-7

APPLICABILITY ACTIONS CS, CRFC and NaOH Systems B3.6.6 In MODES 1, 2, 3, and 4, a OBA could cause a release of radioactive material to containment and an increase in containment pressure and temperature requiring the operation of the CS System, CRFC System and NaOH System.

In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Thus, the CS System, CRFC System and NaOH System are not required to be OPERABLE in MODES 5 and 6.

A.1 With one CS train inoperable, the inoperable CS train must be restored to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or in accordance with the Risk Informed Completion Time Program. In this Condition, the remaining OPERABLE spray and CRFC units are adequate to perform the iodine removal and containment cooling functions. The Completion Times take into account the redundant heat and iodine removal capability afforded by the CRFCs, reasonable time for repairs, and low probability of a OBA occurring during this period.

B.1 With the NaOH System inoperable, OPERABLE status must be restored within 14 days72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The pH adjustment of the Containment Spray System flow for corrosion protection and iodine removal enhancement is reduced in this condition. The Containment Spray System would still be available and would remove some iodine from the containment atmosphere in the event of a OBA. The 14 day72 hour completion time takes into account the redundant flow path capabilities and the low probability of the worst case OBA occurring during this period.

C.1 and C.2 If the inoperable CS train or the NaOH System cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. Tcachievethis status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months . The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. The extended interval to reach MODE 5 allows additional time for attempting restoration of the inoperable component(s) and is reasonable when considering the driving force for a release of radioactive material from the Reactor Coolant System is reduced in MODE 3.

RE. Ginna Nuclear Power Plant B3.6.6-7 Revision 4G7

ATTACHMENT 4 Evaluation of Proposed Editorial Changes

Subject:

License Amendment Request to Extend the Completion Time for Sodium Hydroxide System 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

3.0 REGULATORY EVALUATION

3.1 Applicable Regulatory Requirements 3.2 No Significant Hazards Consideration 3.3 Conclusions

4.0 ENVIRONMENTAL CONSIDERATION

5.0 REFERENCES

1.0

SUMMARY

DESCRIPTION Additional editorial Technical Specification (TS) changes are requested. These changes are the fulfillment of a commitment through the addition of a note, removal of notes that are no longer applicable, completing the implementation of previous amendments and formatting.

A note will be added to Surveillance Requirement (SR) 3.4.14.2 consistent with the regulatory commitment documented in a letter from T.L. Harding to U.S. NRC dated March 2, 2022 (reference 1 ). The note will clarify that Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) check valve pairs are allowed to be tested as a single valve. The note will state, "Check valve pairs without test connections to test valves individually shall be tested as a single valve and shall apply acceptance criteria as if the pair is a single valve."

The note included in TSs 3.4.7, 3.4.8, 3.9.4 and 3.9.5 that allowed an alternative means of Residual Heat Removal (RHR) that was valid between April 3, 2020 and June 30, 2020 is no longer valid and will be removed from TS.

The Spent Fuel Pool (SFP) Charcoal Adsorber System will be removed from the first paragraph of TS 5.5.10. The SFP Charcoal Adsorber System was removed from the Ventilation Filter Testing Program (VFTP) however reference to the SFP Charcoal Adsorber was not removed from the first paragraph of TS 5.5.10.

The Surveillance Requirements table containing SR 3. 7.10.1 and 3. 7.10.2 will have a vertical line added between the surveillance column and the frequency column. A horizontal line will be removed between the note and the Surveillance in SR 3.4.13.2. And typographical errors will be corrected in TS 3.1.6, SR 3.2.1.2, and TS 3.2.3.

2.0 DETAILED DESCRIPTION A note stating, "Check valve pairs without test connections to test valves individually shall be tested as a single valve and shall apply acceptance criteria as if the pair is a single valve." will be added to SR 3.4.14.2. SR 3.4.14.2 requires that Ginna verify leakage from each Safety Injection (SI) system hot leg injection line RCS PIV. The seat leakage is measured, analyzed, and compared to permissible leakage rates at a frequency prescribed by the Surveillance Frequency Control Program (SFCP). As approved in reference 2, series pair check valves will be tested in pairs.

Series pair check valves 877 A/878F and 877B/878H do not have the needed test connections to individually test each valve. Plant modifications to include the proper test connections would be costly and would increase personnel radiation exposure. Due to lack of test connections, each series pair of check valves 877 A/878F and 877B/878H form one of the two pressure boundaries required to be tested with the associated MOV forming the second boundary.

NUREG-1482, Revision 2, "Guidelines for lnservice Testing at Nuclear Power Plants," Section 4.1.1, references ASME OM Code 2012 Edition, paragraph ISTC-5223, which allows the testing of two series check valves as a closed unit if the valves do not have provisions for individual testing and the plant safety analysis requires closure of only one of the series valves.

This relief has been granted in the lnservice Testing Program fifth and sixth 10-year intervals and Ginna committed to including a note in SR 3.4.14.2. Ginna provided a regulatory commitment (reference 1 ), to clarify LCO 3.4.14 to address the prior NRC approval of testing specific check valve pairs as a single isolation valve.

The footnote in TS 3.4.7, 3.4.8, 3.9.4, and 3.9.5 for the spring refueling outage in 2020 allowing an alternate means of RHR as approved in reference 3 will be removed. The note allowed an alternate method of RHR if RHR Pump Suction from Loop A Hot Leg MOV-700 failed to open.

The duration of the applicability of the note was April 3, 2020 to June 30, 2020. The note,

"*Beginning April 3, 2020, an alternative means of RHR as approved in Amendment No. 139 may be used until June 30, 2020. No increase in Mode changes will be permitted while utilizing the alternate approved means for RHR." will be removed from TS 3.4.7, 3.4.8, 3.9.4, and 3.9.5 as it no longer applies.

The Spent Fuel Pool (SFP) Charcoal Adsorber System will be removed from the first paragraph of TS 5.5.10. The first sentence of TS 5.5.10 will be amended from:

A program shall be established to implement the following required testing of Engineered Safety Feature filter ventilation systems and the Spent Fuel Pool (SFP) Charcoal Adsorber System.

to A program shall be established to implement the following required testing of Engineered Safety Feature filter ventilation systems.

The SFP Charcoal Adsorber System was removed from the Ventilation Filter Testing Program in Amendment 153, however SFP Charcoal Adsorber was not removed from the opening paragraph of TS 5.5.10. This change was approved in reference 4, but the SFP Charcoal Adsorber System testing requirement was not removed from the opening paragraph of TS 5.5.10.

The Surveillance Requirements table containing SR 3. 7.10.1 and 3. 7.10.2 will have a vertical line added between the surveillance column and the frequency column. This is standard formatting of the Surveillance Requirements tables found throughout Ginna Technical Specifications and in the Writer's Guide for Plant Specific Improved Technical Specifications TSTF-GG-05-01 revision 1.

The horizontal line dividing the row for SR 3.4.13.2 between the Note and the Surveillance will be deleted. The requirements for the use of lines are in the Writer's Guide for Plant Specific Improved Technical Specifications TSTF-GG-05-01 revision 1. This formatting change is consistent with the guidance, include one surveillance per row.

A typographical error in the note in TS 3.1.6 that reads "Note applicable to control banks inserted while performing SR 3.1.4.2." will be corrected to read "Not applicable to control banks inserted while performing SR 3.1.4.2." A typographical error in SR 3.2.1.2 will be corrected in section bit reads "... maximum over over Z... " one "over will be deleted. Finally, a spelling correction will be made to TS 3.2.3, in the LCO, "spcified" will be corrected to "specified."

3.0 REGULATORY EVALUATION

3.1 Applicable Regulatory Requirements 10 CFR 50.36, Technical Specifications requires Safety Limits, Limiting conditions for operation, Surveillance requirements, and administrative controls. The changes proposed are administrative and do not impact the Safety Limits, LCOs, SRs, or administrative controls.

3.2 No Significant Hazards Consideration A note will be added to Surveillance Requirement (SR) 3.4.14.2 consistent with a regulatory commitment. The note will clarify that Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) check valve pairs are allowed to be tested as a single valve.

The note included in TSs 3.4.7, 3.4.8, 3.9.4 and 3.9.5 that allowed an alternative means of Residual Heat Removal (RHR) that was valid between April 3, 2020 and June 30, 2020 is no longer valid and will be removed.

The Spent Fuel Pool (SFP) Charcoal Adsorber System will be removed from the first paragraph of TS 5.5.10. The SFP Charcoal Adsorber System was removed from the Ventilation Filter Testing Program (VFTP) in a previous amendment (reference 4) however SFP Charcoal Adsorber was not removed from the first paragraph of TS 5.5.10.

The Surveillance Requirements table containing SR 3. 7.10.1 and 3. 7.10.2 will have a vertical line added between the surveillance column and the frequency column. A horizontal line will be removed between the note and the Surveillance in SR 3.4.13.2. And typographical errors will be corrected in TS 3.1.6, SR 3.2.1.2, and TS. 3.2.3.

R.E. Ginna Nuclear Power Plant has evaluated whether a significant hazards consideration is involved with the proposed amendment(s) 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 change to add a note clarifying PIV check valves tested in pairs does not involve a significant increase in the probability or consequence of an accident previously evaluated because it does not propose a change in the current testing methodology it will only add a note for clarity.

The proposed change to remove a note that is no longer valid does not involve a significant increase in the probability or consequence of an accident previously evaluated because the note was only valid for the period between April 3, 2020 and June 30, 2020. This proposed change will not affect any aspect of the way the plant is operated.

The proposed change to remove Spent Fuel Pool Charcoal Adsorber System from the opening paragraph of Technical Specification (TS) section 5.5.1 0 Ventilation Filter Testing Program (VFTP) does not involve a significant increase in the probability or consequence of an accident previously evaluated because the Spent Fuel Pool Charcoal Adsorber System was removed from the VFTP in a previously approved amendment documented in reference 4.

The proposed formatting changes to the lines in the SR tables for SR 3. 7.10.1, 3. 7.10.2, and SR 3.4.13.2 and the editorial changes correcting typographical errors in TS 3.1.6, SR 3.2.1.2, and TS 3.2.3 do not involve a significant increase in the probability or consequence of an accident previously evaluated because these changes are editorial and have no impact on the application of TS or the operation or design of the plant.

Therefore, the proposed changes do 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 change to add a note clarifying PIV check valves tested in pairs does not create the possibility of a new or different kind of accident from any accident previously evaluated because it does not propose a change in the current testing methodology, it will only add a note for clarity.

The proposed change to remove a note that is no longer valid does not create the possibility of a new or different kind of accident from any accident previously evaluated because the note is no longer valid. This proposed change will not affect any aspect of how the plant is operated.

The proposed change to remove Spent Fuel Pool Charcoal Adsorber System from the opening paragraph of TS section 5.5.1 0 Ventilation Filter Testing Program (VFTP) does not create the possibility of a new or different kind of accident from any accident previously evaluated because the Spent Fuel Pool Charcoal Adsorber System was removed from the VFTP in a previously approved amendment (reference 4).

The proposed formatting changes to the lines in the SR tables for SR 3. 7.10.1, 3. 7.10.2, and SR 3.4.13.2 and the editorial changes correcting typographical errors in TS 3.1.6, SR 3.2.1.2, and TS 3.2.3 do not create the possibility of a new or different kind of accident from any accident previously evaluated because these changes are editorial and have no impact on the application of TS or the operation or design of the plant.

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 change to add a note clarifying PIV check valves are allowed to be tested in pairs does not involve a significant reduction in a margin of safety because it does not propose a change in the current testing methodology it will only add a note for clarity.

The proposed change to remove a note that is no longer valid does not involve a significant reduction in a margin of safety because the note is no longer valid. This proposed change will not affect any aspect of the way the plant is operated.

The proposed change to remove Spent Fuel Pool Charcoal Adsorber System from the opening paragraph of TS section 5.5.10 VFTP does not involve a significant reduction in a margin of safety because the Spent Fuel Pool Charcoal Adsorber System was removed from the VFTP in a previously approved amendment (reference 4). This change to the TS will not impact how the VFTP is implemented at Ginna.

The proposed formatting changes to the lines in the SR tables for SR 3. 7.10.1, 3. 7.10.2, and SR 3.4.13.2 and the editorial changes correcting typographical errors in TS 3.1.6, SR 3.2.1.2, and TS 3.2.3 do not involve a significant reduction in a margin of safety because these changes are editorial and have no impact on the application of TS or the operation or design of the plant.

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

Based on the above, CEG concludes that the proposed amendments do not involve a 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.

3.3 Conclusions 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 Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

4.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

5.0 REFERENCES

1. Letter from T.L. Harding to U.S. NRC, "Commitment Associated with technical Specification LCO 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage"' dated March 2, 2022, ML100690308

2. Letter from J. Danna (NRC) to B. Hanson (Exelon Generation Company LLC), "R E. Ginna Nuclear Power Plant-Issuance of Relief Request Associated with Alternatives GR-01, VR-01, and VR-02 for the Sixth 10-Year lnservice Testing Program", dated August 5, 2019, ML19205A353.

3. Letter from V. Sreenivas (NRC) to B. Hanson (Exelon Generation Company LLC), "RE.

Ginna Nuclear Power Plant - Issuance of Amendment No. 139 RE: Add a One-Time Note for Use of Alternative Residual Heat Removal Method", dated April 3, 2020, ML20057E091.

4. Letter from V. Sreenivas (NRC) to D. Rhoades (Constellation Energy Generation, LLC),

"RE. Ginna Nuclear Power Plant-Issuance of Amendment No. 153 RE: Revise Technical Specifications for the Spent Fuel Pool Charcoal System and Two TS Administrative Changes", dated February 23, 2023, ML23005A176.

Markup of Technical Specifications Pages for Editorial Changes R.E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 REVISED TECHNICAL SPECIFICATIONS PAGES 3.1.6-1 3.2.1-4 3.2.3-1 3.4.7-1 3.4.8-1 3.4.13-2 3.4.14-3 3.7.10-1 3.9.4-1 3.9.5-1 5.5-7

I 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Control Bank Insertion Limits Control Bank Insertion Limits 3.1.6 LCO 3.1.6 Control banks shall be within the insertion, sequence, and overlap limits specified in the COLR.

~~~--------------------------------------

-NOTE-Note applicable to control banks inserted while performing SR 3.1.4.2.

APPLICABILITY:

MODE 1, MODE 2 with ketr ~ 1.0.

ACTIONS CONDITION A

Control bank A, B, or A.1 C inserted ~ 8 steps beyond the insertion, sequence,oroverlap limits specified in the COLR.

REQUIRED ACTION Verify the shutdown bank is within the insertion limit specified in the COLR.

COMPLETION TIME 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months A.2.1 Verify SOM is within the limits 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months specified in the COLR.

OR A.2.2 Initiate boration to restore SOM to within limit.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months A.3 Restore the control bank to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months within the insertion, sequence, and overlap limits specified in the COLR.

R.E. Ginna Nuclear Power Plant 3.1.6-1 Amendment ~

( )

( ):

( )

SR 3.2.1.2 SURVEILLANCE

- NOTE-If FQ Z measurements indicate that either the

[ F.c z ]

maximum over z i z or

[ F.w z]

maximum over z Z z

has increased since the previous evaluation of F8 z or if FQ z is expected to increase prior to the next evaluation F8 Z

a.

Increase FQ z by the appropriate factor specified in the COLR and reverify FQ z is within limits specified in the COLR; or

b.

Repeat SR 3.2.1.2 once per 7 EFPD until either

a. above is met or two successive flux maps indicate that the

[ F.c z ]

maximum EWeF over Z i z and

[ F.w z]

maximum over Z Z z has not increased.

Verify FQ Z is within limit.

RE. Ginna Nuclear Power Plant 3.2.1-4 Fo(Z) 3.2.1 FREQUENCY Once after each refueling prior to THERMAL POWER exceeding 75% RTP (continued)

Amendment 448

AFD 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 AXIAL FLUX DIFFERENCE (AFD)

LCO 3.2.3 The AFD in % flux difference units shall be maintained within the limits spsified in the COLR.

!specified V. _______________________________________.

APPLICABILITY:

ACTIONS

- NOTE-The AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.

MODE 1 with THERMAL POWER~ 50% RTP.

CONDITION REQUIRED ACTION COMPLETION TIME A.

AFD not within limits.

A.1 Reduce THERMAL POWER 30 minutes to< 50% RTP.

SURVEILLANCE REQUIREMENTS SR 3.2.3.1 SURVEILLANCE FREQUENCY Verify AFD within limits for each OPERABLE excore In accordance with channel.

the Surveillance Frequency Control Program RE. Ginna Nuclear Power Plant 3.2.3-1 Amendment No. m

RCS Loops - MODE 5, Loops Filled 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4. 7 RCS Loops - MODE 5, Loops Filled LCO 3.4.7 APPLICABILITY:

One residual heat removal (RHR) loop shall be OPERABLE and in operation, and either/ '

a.

One additional RHR loop shall be OPERABLE( or

b.

The secondary side water level of at least one steam generator (SG) shall be ~ 16%.

- NOTE-

1.

The RHR pump of the loop in operation may be de-energized for

~ 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided:

a.

No operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet the SOM of LCO 3.1.1; and

b.

Core outlet temperature is maintained at least 10°F below saturation temperature.

2.

One required RHR loop may be inoperable for~ 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided that the other RHR loop is OPERABLE and in operation.

3.

No reactor coolant pump shall be started with one or more RCS cold leg temperatures less than or equal to the L TOP enable temperature specified in the PTLR unless:

a.

The secondary side water temperature of each SG is ~ 50°F above each of the RCS cold leg temperatures; or

b.

The pressurizer water volume is < 324 cubic feet (38% level).

4.

All RHR loops may be removed from operation during planned heatup to MODE 4 when at least one RCS loop is in operation.

MODE 5 with RCS loops filled.

BegiAAiAg April a, 2020, QA alterAati¥e FAeaAS of RHR as appro*.ied iA AFAeAdFAeAt No. 139 FAay be used UAtil duAe ao, 2020. No iAsrease iA Mode shaAges will be perFAitted while utilii!iAg the altemate appro¥ed FAeaAS for RH R.

RE. Ginna Nuclear Power Plant 3.4.7-1 Amendment 118, 1 ag

RCS Loops - MODE 5, Loops Not Filled 3.4.8 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.8 RCS Loops - MODE 5, Loops Not Filled LCO 3.4.8 Two residual heat removal (RHR) loops shall be OPERABLE and one RHR loop shall be in operation./

APPLICABILITY:

ACTIONS CONDITION A.

One RHR loop inoperable.

- NOTE-

1.

All RHR pumps may be de-energized for s 15 minutes when switching from one loop to another provided:

a.

No operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet the SOM of LCO 3.1.1; and

b.

Core outlet temperature is maintained at least 10°F below saturation temperature; and

c.

No draining operations to further reduce the RCS water volume are permitted.

2.

One RHR loop may be inoperable for s 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided that the other RHR loop is OPERABLE and in operation.

MODE 5 with RCS loops not filled.

REQUIRED ACTION COMPLETION TIME A.1 Initiate action to restore Immediately RHR loop to OPERABLE status.

B.

Both RHR loops B.1 Suspend operations that Immediately inoperable.

would cause introduction of coolant into the RCS with OR boron concentration less than required to meet the No RHR loop in SOM of LCO 3.1.1.

operation.

AND

Be§iRRiR§ /\\13ril a, 2020, QR altematiYe FReRS of RHR as a1313ro1,eel iR l\\FReREIFReRt No. 1 ag FR y be useel uRtil duRe ao, 2020. No iRorease iR Moele oh R§es will be 13erFRitteel while utilii!iR§ the altemate a1313ro1, eel FReRS for RH R.

RE. Ginna Nuclear Power Plant 3.4.8-1 Amendment 118, 1 ag

RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SR 3.4.13.1 SR 3.4.13.2

!Remove line I SURVEILLANCE

- NOTE-

1.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

2.

Not applicable to primary to secondary LEAKAGE.

Verify RCS operational LEAKAGE is within limits by performance of RCS water inventory balance.

- NOTE-Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Verify primary to secondary LEAKAGE is s; 150 gallons per day through any one SG.

RE. Ginna Nuclear Power Plant 3.4.13-2 FREQUENCY In accordance with

~he Surveillance Frequency Control Program In accoraance with

~he Surveillance Frequency Control Program Amendment No. m

SR3.4.14.2 SURVEILLANCE

- NOTE-

1.

Not required to be performed until prior to entering MODE 2 from MODE 3.

2.

RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided.

Verify leakage from each SI hot leg injection line RCS PIV is equivalent to ~ 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm at an RCS pressure~ 2215 psig and~ 2255 psig.

INSERT NOTE

3.

Check valve pairs without test connections to test valves individually shall be tested as a single valve and shall apply acceptance criteria as if the pair is a single valve.

R.E. Ginna Nuclear Power Plant 3.4.14-3 RCS PIV Leakage 3.4.14 FREQUENCY In accordance with he Surveillance Frequency Control Program Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action, flow through the valve, or maintenance on the valve Amendment No. m

3.7 3.7.10 PLANT SYSTEMS Auxiliary Building Ventilation System (ABVS)

LCO3.7.10 The ABVS shall be OPERABLE and in operation.

ABVS 3.7.10 APPLICABILITY:

During movement of irradiated fuel assemblies in the Auxiliary Building when one or more fuel assemblies in the Auxiliary Building has decayed < 60 days since being irradiated.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

ABVS inoperable.

A.1

-NOTE-LCO 3.0.3 is not applicable.

Suspend movement of Immediately irradiated fuel assemblies in the Auxiliary Building.

SURVEILLANCE REQUIREMENTS SR3.7.10.1 SR3.7.10.2 SURVEILLANCE Verify ABVS is in operation.

Verify ABVS maintains a negative pressure with respect to the outside environment at the Auxiliary Building operating floor level.

R.E. Ginna Nuclear Power Plant 3.7.10-1 pnsert line. I I

I FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment No. 4-aa

RHR and Coolant Circulation - Water Level 2 23 Ft 3.9.4 3.9 REFUELING OPERATIONS 3.9.4 Residual Heat Removal (RHR) and Coolant Circulation - Water Level 2 23 Ft LCO 3.9.4 One RHR loop shall be OPERABLE and in operation/

APPLICABILITY:

ACTIONS CONDITION

-NOTE-The required RHR loop may be removed from operation for~ 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period, provided no operations are permitted that would cause introduction of coolant into the Reactor Coolant System (RCS) with boron concentration less than that required to meet the minimum required boron concentration of LCO 3.9.1.

MODE 6 with the water level 2 23 ft above the top of reactor vessel flange.

REQUIRED ACTION COMPLETION TIME A.

RHR loop requirements A.1 Suspend operations that Immediately not met.

would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.

AND A.2 Suspend loading irradiated Immediately fuel assemblies in the core.

AND A.3 Initiate action to satisfy RHR Immediately loop requirements.

AND

Be§iRRiR§ April a, 2020, aR allerRafr,e FReaRS of RHR as approYeel iR AFReRelFReRl ~lo. 139 FRay be useel uRtil duRe ao, 2020. No iRorease iR Moele ohaRges 1uill be perFRitteel 1uhile utilii!iRg the allerRate appro1,eel FReaRS for RHR.

RE. Ginna Nuclear Power Plant 3.9.4-1 Amendment 118, 1 ag

3.9 REFUELING OPERATIONS RHR and Coolant Circulation - Water Level < 23 Ft 3.9.5 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation - Water Level < 23 Ft LCO 3.9.5 Two RHR loops shall be OPERABLE, and one RHR loop shall be in operation/

APPLICABILITY:

MODE 6 with the water level < 23 ft above the top of reactor vessel flange.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Less than the required A.1 Initiate action to restore Immediately number of RHR loops RHR loop(s) to OPERABLE OPERABLE.

status.

OR A.2 Initiate action to establish Immediately 2 23 ft of water above the top of reactor vessel flange.

B.

No RHR loop in 8.1 Suspend operations that Immediately operation.

would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.

AND 8.2 Initiate action to restore one Immediately RHR loop to operation.

AND 8.3 Close all containment 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months penetrations providing direct access from containment to outside atmosphere.

Be§iAAiA§ April a, 2020, QA alleFAafr,e FAeOAS of RHR as approved iA AFAeAdFAeAl No. 139 FAay be used uAlil duAe ao, 2020. No iAorease iA Mode ohaA§es 1h'ill be perFAitted while ulilii!:iA§ the alleFAale approved FAeOAS for RHR.

R.E. Ginna Nuclear Power Plant 3.9.5-1 Amendment 118, 1 ag

5.5.9 5.5.10 Secondary Water Chemistry Program Programs and Manuals 5.5 This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. This program shall include:

a.

Identification of a sampling schedule for the critical variables and control points for these variables;

b.

Identification of the procedures used to measure the values of the critical variables;

c.

Identification of process sampling points;

d.

Procedures for the recording and management of data;

e.

Procedures defining corrective actions for all off control point chemistry conditions; and

f.

A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.

Ventilation Filter Testing Program NFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature filter ventilation systems and tl:ie S13ent Fuel Pool (SFP) GhaFooal /\\dsoFl3eF SysteFA. The test frequencies will be in accordance with Regulatory Guide 1.52, Revision 2, except that in lieu of 18 month test intervals, a 24 month interval will be implemented.

The test methods will be in accordance with Regulatory Guide 1.52, Revision 2, except as modified below.

a.

Containment Recirculation Fan Cooler System

1.

Demonstrate the pressure drop across the high efficiency particulate air (HEPA) filter bank is < 3 inches of water at a design flow rate(+/- 10%).

2.

Demonstrate that an in-place dioctylphthalate (DOP) test of the HEPA filter bank shows a penetration and system bypass

< 1.0%.

b.

Control Room Emergency Air Treatment System (CREA TS)

1.

Demonstrate the pressure drop across the combined HEPA filters, the prefilters, the charcoal adsorbers and the post-filters is< 11 inches of water at a design flow rate (+/- 10%).

RE. Ginna Nuclear Power Plant 5.5-7 Amendment 44G

Markup of Technical Specification Bases Pages for Editorial Changes R.E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 REVISED TECHNICAL SPECIFICATION BASES PAGES B 3.1.6-4 B 3.4.14-6 B 3.4.14-7

LC0 APPLICABILITY Control Bank Insertion Limits 83.1.6 Operation at the insertion limits or AFD limits may approach the maximum allowable linear heat generation rate or peaking factor with the allowed QPTR present. Operation at the insertion limit may also indicate the maximum ejected RCCA worth could be equal to the limiting value in fuel cycles that have sufficiently high ejected RCCA worths.

The control and shutdown bank insertion limits, together with AFD, QPTR and the control and shutdown bank alignment limits, ensure that safety analyses assumptions for SOM, ejected rod worth, and power distribution peaking factors are preserved (Refs. 4, 5, 6, and 7).

The control bank insertion, sequence and overlap limits satisfy Criterion 2 of the NRC Policy Statement, in that they are initial conditions assumed in the safety analysis.

The limits on control bank insertion, sequence, overlap, and as specified in the COLR, must be maintained because they serve the function of preserving power distribution, ensuring that the SOM is maintained, ensuring that ejected rod worth is limited, and ensuring adequate negative reactivity insertion is available on trip. The overlap between control banks provides more uniform rates of reactivity insertion and withdrawal and is imposed to maintain acceptable power peaking during control bank motion.

The LCO is modified by a Note indicating the LCO requirement is not applicable to control banks being inserted while performing SR 3.1.4.2.

This SR verifies the freedom of the rods to move, and may require the control bank to move below the LCO limits, which would normally violate the LCO. This Note applies to each control bank as it is moved below the insertion limit to perform the SR. This Note applies to eaoh oontrol bani< as it is mo¥ed below the insertion limit to perform the SR. This Note is not applicable should a malfunction stop performance of the SR.

The control bank insertion, sequence, and overlap limits shall be maintained with the reactor in MODE 1 and MODE 2 with keff 2:1.0.

These limits must be maintained, since they preserve the assumed power distribution, ejected rod worth, SOM, and reactivity rate insertion assumptions. Applicability in MODE 2 with keff < 1.0 and MODES 3, 4, 5, and 6 is not required, since neither the power distribution nor ejected rod worth assumptions would be exceeded in these MODES.

RE. Ginna Nuclear Power Plant 83.1.6-4 Revision 88

The SR is modified by three notes. The third note clarifies the allowance the NRC has previously approved to test check valve pairs as a single valve.

Each check valve pair 877A/878F and 877B/878H shall be tested as a single valve with the acceptance criteria as if it were a single valve. (commitment per reference 12)

The SI hot leg injection lines are each configured with two check valves and a motor operated valve in series. Each of these components independently is considered a qualified pressure boundary. The two check valves function as a single pressure isolation barrier and the motor operated valve serves as the second pressure isolation barrier to prevent an intersystem LOCA. Both barriers need to be tested. Testing of the check valves (877A, 877B, 878F, and 878H) and the motor operated valves (878A and 878C) identified as PIVs in the SI hot leg injection lines is to be performed in accordance with the Surveillance Frequency Control Program.

RCS PIV Leakage B 3.4.14 The SI hot leg injestion lines are eaoh oonfigured with two oheol< Yalves and a rnotor operated YalYe in series. Eaoh of these oornponents independently is oonsidered a qualified pressure boundary. The two oheGI< valves funGtion as a single pressure isolation barrier and the rnotor operated valve serves as the seoond pressure isolation barrier to pre1,ent an intersystern LOGA. Both barriers need to be tested. Testing of the oheol< Yalves (877A, 877B, 878F, and 8781=1) and the rnotor operated valYes (878A and 878G) identified as PIVs in the SI hot leg injestion lines is to be perforrned in aooordanoe with the Surveillanoe Frequenoy Control Prograrn.

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

In addition to the periodic testing requirements, testing must be performed once after the valve has been opened by flow, exercised, or had maintenance performed on it to ensure tight reseating. This maintenance does not include minor activities such as packing adjustments which do not affect the leak tightness of the valve. PIVs disturbed in the performance of this Surveillance should also be tested unless documentation shows that an infinite testing loop cannot practically be avoided. Testing must be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the valve has been reseated. A limit of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable and practical time limit for performing this test after opening or reseating a valve.

The leakage limit is to be met at the RCS pressure associated with MODES 1 and 2. This permits leakage testing at high differential pressures with stable conditions not possible in the MODES with lower pressures.

Entry into MODES 3 and 4 is allowed to establish the necessary differential pressures and stable conditions to allow for performance of this Surveillance.

SR3.4.14.2 R.E. Ginna Nuclear Power Plant B 3.4.14-6 Revision

12. Letter to USNRC Document Control Desk from Thomas Harding (Ginna), Commitment Associated with Technical Specification LCO 3.4.14 RCS Pressure Isolation Valve (PIV)

Leakage, Dated March 2, 2010 REFERENCES

1.

10CFR50.2.

2.

10 CFR 50.55a( c).

RCS PIV Leakage B 3.4.14

3.

Atomic Industry Forum (AIF) GDC 53, Issued for comment July 10, 1967.

4.

WASH-1400 (NUREG-75/014), "An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," Appendix V, October 1975.

5.

NUREG-0677, "The Probability of lntersystem LOCA: Impact Due to Leak Testing and Operational Changes," May 1980.

6.

Generic Letter, "LWR Primary Coolant System Pressure Isolation Valves," dated February 23, 1980.

7.

Deleted.

8.

EG&G Report, EGG-NTAP-6175.

9.

Deleted.

10.

Deleted.

11.

Deleted.

R.E. Ginna Nuclear Power Plant B 3.4.14-7 Revision +l-

v t e Letter Sequence Request
EPID:L-2025-LLA-0087, (Open)
Initiation Acceptance... Supplement Administration Meeting Questions Answers Results Approval... Other: ML25163A306, ML25233A245
MONTHYEAR2025-06-12 [Table View]

ML25140A128

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Subject:

Subject:

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

Subject:

SUMMARY

3.0 REGULATORY EVALUATION

4.0 ENVIRONMENTAL CONSIDERATION

5.0 REFERENCES

SUMMARY

3.0 REGULATORY EVALUATION

4.0 ENVIRONMENTAL CONSIDERATION

5.0 REFERENCES

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