Text

Emergency License Amendment Request for Technical Specification 3.7.2 Regarding One-Time Extension of Completion Time for Plant Service Water (Psw) Pump Inoperable
	ML22120A087
Person / Time
Site:	Hatch
Issue date:	11/30/2022
From:	Gayheart C; Southern Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NL-22-0316
	Download: ML22120A087 (50)
	v • d • e

Cheryl A. Gayheart 3535 Colonnade Parkway Regulatory Affairs Director Birmingham, AL 35243 205 992 5316

cagayhea@southernco.com

April 30, 2022

Docket No.: 50-321 NL-22-0316

U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001

Edwin I. Hatch Nuclear Plant Unit 1 Emergency License Amendment Request for Technical Specification 3.7.2 Regarding One-Time Extension of Completion Time for Plant Service Water (PSW) Pump Inoperable Ladies and Gentlemen:

Pursuant to the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), Southern Nuclear Operating Company (SNC) hereby requests a license amendment to the Technical Specifications (TS) for Hatch Nuclear Plant (HNP) Unit 1 renewed facility operating license DPR-57.

The proposed amendment would revise TS 3.7.2, Plant Service Water (PSW) System and Ultimate Heat Sink (UHS), Limiting Condition for Operation (LCO) 3.7.2, Required Actions for Condition A, One PSW pump inoperable, to allow a one-time increase in the Completion Time from 30 days to 45 days. The extended Completion Time would expire on May 19, 2022, at 1527 eastern daylight time (EDT).

The one-time only change allows for continued repair and testing activities on the HNP Unit 1 1A PSW Pump. The expiration date for the proposed allowance is based on the current 30-day Completion Time expiration at 1527 EDT, May 4, 2022, plus the requested additional 15 days (45 days total).

This proposed amendment to the HNP Unit 1 TS is being requested on an emergency basis for the Unit 1 PSW System, pursuant to 10 CFR 50.91(a)(5). The Unit 2 PSW System is not affected by this proposed amendment.

SNC requests approval of the proposed license amendment as soon as possible and no later than May 3, 2022, based on emergent circumstances at HNP Unit 1 in accordance with the provisions of 10 CFR 50.91(a)(5). A discussion of the emergency situation is provided in the enclosure to this letter. The amendment, if approved, will be implemented immediately upon issuance.

The Enclosure provides a description and assessment of the proposed change, including a no significant hazards considerations analysis, regulatory requirements, and environmental considerations. Attachments 1 and 2 contain a marked-up TS page and revised TS page, U.S. Nuclear Regulatory Commission NL-22-0316 Page 2 of 2

respectively, reflecting the proposed changes. Attachment 3 contains a markup of the TS Bases, for information only. Attachment 4 contains an evaluation of the risk impact and a discussion of the compensatory measures related to the changes in this license amendment request.

In accordance with the SNC administrative procedures and the HNP quality assurance program manual, this proposed license amendment has been reviewed by the Plant Review Board and approved by the Plant Manager.

In accordance with 10 CFR 50.91, SNC is notifying the State of Georgia of this license amendment request by transmitting a copy of this letter, enclosure, and attachments to the designated State Official.

This letter contains no regulatory commitments. If you have any questions, please contact Ryan Joyce at 205-992-6468.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 30th day of April 2022.

Respectfully submitted,

Cheryl A. Gayheart Regulatory Affairs Director

CAG/tle

Enclosure:

Description and Assessment of the Proposed Change Attachments:

1. HNP Unit 1 Technical Specification Marked-up Page

2. HNP Unit 1 Revised Technical Specification Page

3. HNP Unit 1 Technical Specification Bases Marked-up Page (information only)

4. Evaluation of Risk Impact and Compensatory Measures

cc: NRC Regional Administrator, Region II NRC NRR Project Manager - Hatch NRC Senior Resident Inspector - Hatch Director, Environmental Protection Division - State of Georgia RType: CHA02.004

Edwin I. Hatch Nuclear Plant Unit 1

Emergency License Amendment Request for Technical Specification 3.7.2 Regarding One-Time Extension of Completion Time for Plant Service Water (PSW) Pump Inoperable

Enclosure

Description and Assessment of the Proposed Change

Enclosure to NL-22-0316 Description and Assessment of the Proposed Change

1. Summary Description The proposed amendment to Hatch Nuclear Plant (HNP) Unit 1 renewed facility operating license DPR-57 would revise Technical Specification (TS) 3.7.2, Plant Service Water (PSW) System and Ultimate Heat Sink (UHS), Limiting Condition for Operation (LCO) 3.7.2, Required Actions for Condition A, One PSW pump inoperable, to allow a one-time increase in the Completion Time from 30 days to 45 days. The allowance would also require implementation of compensatory measures described in Section 3.3 of this application and as affirmed in the NRCs safety evaluation, would only apply to the 1A PSW Pump, and would expire on May 19, 2022, at 1527 eastern daylight time (EDT).

On April 4, 2022, at 1527 EDT, the HNP 1A PSW Pump was tagged out and declared inoperable to perform a preventive maintenance task. As described in Section 2.1 of this application, the original scope of work was to replace the motor which was anticipated to take four to five days. Delays due to motor alignment led to placing the pump in service on April 10, 2022. During the first run after motor replacement, vibrations steadily rose to the point that the pump had to be removed from service. The motor and pump were removed and shipped to their respective vendors. It was determined that there were no problems with the motor. During the pump inspection, the discharge head was found to have measurements that were out of tolerance which likely resulted in misalignment when the motor was coupled to the pump. To correct this issue, the discharge head was machined and attached to a new pump assembly. This assembly was sent back to the site and installed on April 25, 2022. On April 26, 2022, while attempting to bolt the discharge head to the discharge check valve, it was determined that the discharge check valve flange was too far out of alignment with the pump discharge head flange, thus requiring work to the discharge check valve flange. As discussed in Section 2.1, additional discovery on April 26, 2022, has required HNP to remove the pump and rework the flange on the discharge check valve off-site at a specialty vendor.

The comprehensive repair work has been time-consuming; however, Southern Nuclear Operating Company (SNC) has demonstrated due diligence by safely performing testing and maintenance activities without delay. SNC now anticipates that repair and testing will extend past the 30-day Completion Time of the above-listed TS and therefore requests additional time to make careful, prudent repairs with appropriate compensatory measures in place to return the HNP 1A PSW Pump to operable status. To provide allowance for additional time to complete repairs, SNC is requesting a one-time Completion Time extension from 30 days to 45 days, applicable to the 1A PSW Pump only.

2. Detailed Description 2.1 Emergency Circumstances On April 4, 2022, at 1527 EDT, HNP 1A PSW Pump was tagged out and declared inoperable to perform a preventive maintenance task, specifically replacement of the pump motor. On April 10, 2022, during the first run after motor replacement, vibration levels rose to the point that the pump had to be removed from service and replaced. After new pump and motor replacement on April 26, 2022, significant

E-1 Enclosure to NL-22-0316 Description and Assessment of the Proposed Change

misalignment was discovered between the pump discharge flange and the discharge piping. The emergency circumstances result from this additional discovery on April 26, 2022, eight days before the expiration of the required Completion Time for TS 3.7.2 Required Action A.1. The events that led to these emergency circumstances are documented below.

Why the Condition Occurred:

On April 10, 2022, at 1846 EDT, HNP 1A PSW Pump was secured due to excessive vibration as measured at the motor. Prior to motor replacement, the pump and motor had shown good performance with respect to vibration, with the previous four In-Service Tests (ISTs) being satisfactory. The emergency circumstances of this LAR result from the unanticipated extended repairs. The extension of this window is a result of significant misalignment discovered between the pump discharge flange and the discharge piping on April 26, 2022, eight days before the expiration of the Completion Time for Required Action A.1 for TS 3.7.2. The events that led to these emergency circumstances and the facts surrounding the initial event (Initial Discovery) and misalignment (Additional Discovery) are documented below. SNC has been performing test and repair activities, and the pump and motor vendors are engaged with the investigation and repair activities. The TS 3.7.2 Required Action A.1 Completion Time of 30 days is currently applicable and will expire on May 4, 2022, at 1527 EDT. The current repair and replacement activities along with the identified contingencies associated with repairs are unlikely to be completed during the remaining allowed Completion Time. Neither a routine nor an exigent amendment can be processed prior to May 4, 2022, at 1527 EDT.

It is noted that at the time of Initial Discovery up to Additional Discovery, SNC had reasonable expectation that the TS Completion Time would be met.

Initial Discovery On April 4, 2022, at 1527 EDT, the 1A PSW Pump was tagged out for scheduled preventive maintenance to replace the motor. After placing the pump in service on April 10, 2022, at 0151 hours0.00175 days 0.0419 hours 2.496693e-4 weeks 5.74555e-5 months , the 1A PSW Pump motor experienced high vibration levels. Vibration levels on the 1A PSW Pump motor were initially taken at 0240 hours0.00278 days 0.0667 hours 3.968254e-4 weeks 9.132e-5 months on April 10, 2022, and maximum vibration was 0.402 in/sec.

Based on visible vibration observations later during dayshift on April 10, vibration levels were obtained again at 1755 hours0.0203 days 0.488 hours 0.0029 weeks 6.677775e-4 months . All vibration levels had increased with the horizontal vibration at the top of the motor increasing to 1.318 in/sec which placed it well above the Unacceptable limit of 0.750 in/sec. Operations therefore secured the 1A PSW Pump motor at 1846 hours0.0214 days 0.513 hours 0.00305 weeks 7.02403e-4 months based on these vibration readings. The motor was sent to a vendor to perform operational assessments and diagnostics to determine if the motor had been damaged. The results showed the motor had not sustained any damage.

With uncertainties surrounding damage to the pump, SNC made the decision to replace the pump. This removed any new discovery points associated with potential pump damage while ensuring a reliable 1A PSW Pump.

The replacement of both the pump and motor were on schedule within the bounds of the 30-day Completion Time. The replacement pump was found to have measurements on

E-2 Enclosure to NL-22-0316 Description and Assessment of the Proposed Change

the discharge head that were out of tolerance which resulted in misalignment when the motor was installed on the pump. To correct this issue, the discharge head was machined and attached to the new pump assembly. This assembly was sent back to the site and installed on April 25, 2022.

Additional Discovery On April 26, 2022, during attempts to bolt the pump discharge head to the discharge check valve, it was determined that the discharge check valve flange was too far out of alignment with the pump discharge head flange, thus requiring modification work to align the discharge components. This additional discovery has required SNC to remove the pump to access the discharge check valve, which has been sent off-site to a specialty vendor for modification of its flange to address the alignment challenges. As shown in Section 2.4, this additional work leads SNC to conclude it is unlikely that the 1A PSW Pump can be returned to operable service by May 4, 2022, 1527 EDT.

Why this Situation Could Not be Avoided:

Based on the investigation performed following the initial event on April 10, 2022, with a return to service date of April 26, 2022, at 2300 EDT, SNC had sufficient margin to complete the repairs prior to the expiration of the TS Completion Time. Due to the subsequent unexpected discovery with the pump discharge piping misalignment, the expected replacement duration was delayed.

Need for an Emergency Amendment pursuant to 10 CFR 50.91(a)(5)

Immediately following the post-maintenance test run on April 10, 2022, SNC began investigating the cause and conducting repairs of the post-maintenance test vibration issue. It was unknown at that time that additional misalignment issues would be discovered on April 26, 2022, that would challenge the TS allowed Completion Time. On April 27, 2022, seven days prior to the expiration of the applicable TS Completion Time, it was concluded that the pump and motor replacement, and subsequent operability testing to restore compliance with TS 3.7.2, would not be completed prior to expiration of the current TS Completion Time due to complications with alignment of the discharge piping. Without delay, work commenced to develop the schedule and secure the needed technical expertise to complete these repairs.

The determination of the requested extension duration to the TS Completion Time involved a thorough review of the repair schedule to assure: (a) the request was reasonable (i.e., not excessive and unnecessarily long); and (b) there is high probability that the 1A PSW Pump would be restored to operable status within the requested Completion Time extension.

In addition, to provide assurance that the incurred risk to the plant is acceptable and that high-risk configurations will be avoided during the extended TS Completion Time, a risk analysis meeting the requirements of Regulatory Guide 1.177, An Approach for Plant-Specific, Risk-Informed Decision Making: Technical Specifications, Revision 2 (Reference 1) was performed. This analysis is summarized in Attachment 4 of this letter.

This risk analysis identified compensatory m easures, described in Section 3.3 of this

E-3 Enclosure to NL-22-0316 Description and Assessment of the Proposed Change

enclosure, that are required to be implemented during the duration of the proposed extended Completion Time.

SNC has been diligent in the preparation of this LAR. Processing an exigent LAR pursuant to 10 CFR 50.91(a)(6) requires reasonable notice to the public in the areas surrounding the facility of the proposed LAR and of the determination of the no significant hazards consideration as described in 10 CFR 50.91(a)(2). However, based on both: (a) the timing of the realization that a LAR is needed to avoid a TS required shutdown; and (b) the development of required inputs to prepare this LAR (e.g., detailed repair schedule), submittal of an exigent LAR was not feasible. Therefore, an emergency amendment pursuant to 10 CFR 50.91(a)(5) is necessary.

To allow adequate time for the preparation of a TS required shutdown (in the event this amendment is denied), SNC is requesting NRC approval by May 3, 2022 (i.e., six days after it was determined an emergency TS change was required). Therefore, an exigent TS change per 10 CFR 50.91(a)(6) was not feasible once the full scope of the pump and motor repair was understood.

Operational Experience (OE) Review SNC has reviewed recent applicable operating experience within our fleet and the industry regarding vertical pumps. The most relevant OE came from an HNP Plant Service Water pump (1C PSW Pump) that experienced a failed pump shaft due to intergranular stress corrosion cracking that was exacerbated by alignment issues with the discharge head and the stuffing box. The situation was corrected by replacing the pump.

Additional OE from SNCs Farley Nuclear Plant noted damage to a vertical pump due to a wiring issue. The 1A PSW Pump does not show any indications of having wiring related failures.

In addition to these specific SNC OE reviews, various industry and fleet subject matter experts have been consulted throughout this system outage to ensure that broader knowledge of potential issues is incorporated into the review and return to service of the 1A PSW Pump. The inclusion of the feedback from subject matter experts, as well as validation of other fleet OE noted above, is being utilized to ensure a full understanding and resolution of any anomalies prior to return to service.

In addition, industry and pump vendor operating experience (OE) was reviewed, as well as NRC Information Notices (INs) (e.g., INs 93-68, 94-45, 96-36, and 07-05). This review did not identify any OE that would suggest a potential common cause mode failure for the 1A PSW Pump.

Common Cause Consideration The 1A PSW Pump was removed from service for preventive maintenance. As such, the cause of the inoperability is known and does not subject the other PSW pumps to potential common cause. The discovered alignment issue is not common to the other Unit 1 pumps since it occurred during intrusive maintenance where new mechanical components were being assembled. It is also noted that vibration data, oil quality, and

E-4 Enclosure to NL-22-0316 Description and Assessment of the Proposed Change

bearing temperature of each pump is peri odically taken and demonstrates that the remaining PSW pumps are operating without indication of degradation.

Summary of Emergency Circumstances:

In summary, the emergency circumstances resulted from unforeseen alignment issues of the HNP 1A PSW Pump first identified during post-repair installation of the pump and motor on April 26, 2022. The required Completion Time for TS 3.7.2 Required Action A.1 of 30 days is currently applicable and will expire on May 4, 2022, at 1527 EDT. SNC is unlikely to complete repairs of the 1A PSW Pump prior to this time.

Neither a routine nor an exigent amendment can be processed prior to May 4, 2022, at 1527 EDT.

SNC has performed due diligence by safely performing testing and maintenance activities around the clock. The repair plan included many provisions to ensure timely execution of the work including the use of experienced personnel, pre-assembled components, and pre-staging of equipment. Experienced pump and motor vendors are engaged for assistance. Therefore, efforts were made to minimize the likelihood for delays due to job planning or preparation.

SNC requests an expedited review of the proposed license amendment in accordance with the provisions of 10 CFR 50.91(a)(5) based on avoiding the need to shut down HNP Unit 1 without an approved amendment. If the proposed license amendment is not approved, Unit 1 will be required to enter TS 3.7.2 Condition E on May 4, 2022, at 1527 EDT, which requires a plant shutdown.

Based on the discussion herein, SNC has determined that emergency circumstances exist, has used its best efforts to make a timely application, and did not knowingly cause the emergent situation.

2.2 System Design and Operation The PSW System is designed to provide cooling water for the removal of heat from equipment, such as the diesel generators (DGs), residual heat removal (RHR) pump coolers, and room coolers for Emergency Core Cooling System equipment, required for a safe reactor shutdown following a Design Basis Accident (DBA) or transient. The PSW System also provides cooling to unit components, as required, during normal operation. Upon receipt of a loss of offsite power or loss of coolant accident (LOCA) signal, nonessential loads are automatically isolated, the essential loads are automatically divided between PSW Divisions 1 and 2, and one PSW pump is automatically started in each division.

The PSW System consists of the ultimate heat sink (UHS) and two independent and redundant subsystems. Each of the two PSW subsystems is made up of a header, two 8500 gpm pumps, a suction source, valves, piping and associated instrumentation.

Either of the two subsystems is capable of providing the required cooling capacity to support the required systems with one pump operating. The two subsystems are separated from each other so failure of one subsystem will not affect the operability of the other system.

E-5 Enclosure to NL-22-0316 Description and Assessment of the Proposed Change

Cooling water is pumped from the UHS (i.e., the Altamaha River) by the PSW pumps to essential components through the two main headers. After removing heat from the components, the water is discharged to the circulating water flume to replace evaporation losses from the circulating water sy stem, or directly to the river via a bypass valve.

The ability of the PSW System to support long term cooling of the reactor containment is assumed in evaluations of the equipment required for safe reactor shutdown presented in the FSAR, Section 10.7 (Reference 2). These analyses include the evaluation of the long-term primary containment response after a design basis LOCA.

The ability to provide onsite emergency AC power is dependent on the ability of the PSW System to cool the DGs. The long-term cooling capability of the RHR, core spray, and RHR service water pumps is also dependent on the cooling provided by the PSW System. In the analysis presented in Reference 2, only one PSW pump is required for safe shutdown, including RHR Shutdown Cooling System requirements.

The PSW subsystems are independent of each other to the degree that each has separate controls and power supplies, and the operation of one does not depend on the other. In the event of a DBA, one PSW pump is required to provide the minimum heat removal capability assumed in the safety analysis for the system to which it supplies cooling water. To ensure this requirement is met, two subsystems, each with two pumps, of PSW must be operable. At least one pump will operate if the worst single active failure occurs coincident with the loss of offsite power.

A subsystem is considered operable when it has an operable UHS, two operable pumps, and an operable flow path capable of taking suction from the intake structure and transferring the water to the appropriate equipment.

2.3 Current Technical Specification Requirements Currently, TS 3.7.2, Required Action A.1, requires restoration of an inoperable PSW pump to operable status within 30 days.

2.4 Reason for the Proposed Change Despite diligent and prudent efforts, SNC has been unable to return the 1A PSW Pump to operable service and now expects it will be unable to do so by expiration of the Completion Time for TS 3.7.2, Required Action A.1 (i.e., May 4, 2022, at 1527 EDT) and would then be required to place HNP Unit 1 in Mode 3 (i.e., Hot Shutdown). Approval of the proposed change would allow SNC to continue repair and replacement activities as necessary and without undue risk as demonstrated in Attachment 4 of this license amendment request.

The table below summarizes a conservative schedule with maintenance tasks assuming additional rework and additional contingencies. SNC will work to restore the 1A PSW Pump to operable status as soon as reasonably and safely achievable, regardless of the additional Completion Time duration.

E-6 Enclosure to NL-22-0316 Description and Assessment of the Proposed Change

Task No. Project Task Expected Task Completion Date

1 Remove Pump for any Required Additional Work 5/2/2022

2 Ship Component(s) Offsite for any Required 5/2/2022 Rework or Repair (or perform onsite)

3 Receive Reworked Components 5/5/2022

4 Install Reworked Components 5/5/2022

5 Install Pump 5/6/2022

6 Install Motor 5/8/2022

7 Restore Pump to Operable 5/9/2022

8 Contingency* for Additional Discoveries and 5/19/2022 Rework as Necessary

Contingency allowance includes potential weather delays and post-installation operability testing (considers the 1A PSW Pump is a new unit with no current run time).

While SNC believes the above schedule is realistic, there are inherent uncertainties and unknowns associated with major pump and motor maintenance and testing. To account for these uncertainties and unknowns, SNC is requesting contingency as addressed by Item 8 of the schedule. Thus, as supported by this LAR, SNC requests until May 19, 2022, at 1527 EDT (i.e., a total 45-day Completion Time) to complete this work.

Approval of the proposed change would allow SNC to continue repair, refurbishment, and replacement activities as necessary and without undue risk as demonstrated in Attachment 4 of this license amendment request.

E-7 Enclosure to NL-22-0316 Description and Assessment of the Proposed Change

2.5 Description of the Proposed Change The following revision is proposed to TS 3.7.2 Required Actions associated with Condition A. This revision modifies text added in HNP Unit 1 Amendment 311 (see Section 4.2). Deleted text is shown in red font with strikethrough; added text is shown in blue font and underlined.

The one-time only change allows for continued repair and testing activities in a prudent fashion. The expiration date of May 19, 2022, at 1527 EDT is based on the current 30-day Completion Time expiration at 1527 EDT, May 4, 2022, plus the requested additional 15 days (for a total Completion time of 45 days). The allowance allows SNC to safely address any additional unforeseen circumstances, such as weather conditions, or discoveries, such as the need to order new parts or to contract with a specialty vendor. The allowance also considers the 1A PSW Pump is a new unit with no current

E-8 Enclosure to NL-22-0316 Description and Assessment of the Proposed Change

run time. The allowance would only apply to the 1A PSW Pump and only as long as the compensatory measures described in Section 3.3 of this application are implemented.

3. Technical Evaluation 3.1 Defense-in-Depth During the extended Completion Time, the PSW System will remain in compliance with the Technical Specifications. Should an event occur requiring the function of the PSW System and the UHS (i.e., the Altamaha River), the remaining PSW pumps are capable of performing the safety function of providing the necessary cooling water.

As an additional defense-in-depth measure, if a total loss of PSW is assumed, the station would implement the core cooling FLEX strategy. For this strategy, the on-site 600V FLEX generators and associated booster pumps would be utilized.

The on-site 600V FLEX generators and the booster pump components used in the core cooling FLEX strategy are currently available, and all maintenance is current on these components. A review of recent Condition Reports did not reveal any reliability issues associated with this equipment. No preventive maintenance activities for these components will be performed during the extended Completion Time.

Only permanently installed FLEX equipment and related actions are modeled in the Seismic PRA model for defense-in-depth purposes. There is no credit for other FLEX equipment or actions in the Seismic PRA model.

In addition, compensatory measures as described in Section 3.3 will be in place and available.

3.2 Safety Margin Evaluation The proposed TS change is consistent with the principle that sufficient safety margins are maintained based on the following:

Codes and standards (e.g., American Society of Mechanical Engineers (ASME),

Institute of Electrical and Electronic Engineers (IEEE), or alternatives approved for use by the NRC) are met. The propos ed change is not in conflict with approved codes and standards relevant to the PSW System.

The PSW system has sufficient capacity to function for design basis events while in TS 3.7.2 Condition A (one PSW pump inoperable). The UFSAR acceptance criteria for the design events will be met should such an event occur during the time that the 1A PSW Pump is out of service. It is noted that in the analysis presented in Reference 2, only one PSW pump is required for safe shutdown, including RHR Shutdown Cooling System requirements.

E-9 Enclosure to NL-22-0316 Description and Assessment of the Proposed Change

3.3 Compensatory Measures The following compensatory measures are required prior to exceeding the original 30-day CT.

The following equipment is protected as required by SNC Procedure NMP-OS-010-002 (Reference 3) for 1A PSW Pump out-of-service:

o 1C PSW Pump o 1F 4160V Frame 4 (power supply to 1C PSW Pump) o 1C PSW Pump Control Switch

Site walkdown to remove loose objects per site procedure 34AB-Y22-002-0 (Reference 4)

Additionally, the 1B and 1D PSW subsystems are being protected by Operations to mitigate the entry into another TS Action (TS 3.7.2 Action C). Plant Hatch Operations have initiated protecting the 1B, 1C, and 1D PSW pumps prior to submittal of this LAR.

Based on the operator actions and equipment above, the following additional compensatory actions will be implemented:

The operator actions contained in procedure 34AB-P41-001-1 (Reference 5) and an Operations Standing Order for shift briefing on the procedure associated with these actions will be in place prior to exceeding the original 30-day CT.

o Section 4.1.7 Crosstie Cooling Sources This ensures that adequate cooling will remain available for the RHRSW pump o Section 4.1.9 Operating Strategy for Event Mitigation Preparation for potential loss of an additional PSW pump

Protect the 1C diesel generator, 1B diesel generator, and the Standby Service Water (SSW) pump 2P41C002 that provides normal cooling to the 1B DG and limit work to TS required TS surveillances only.

Work in the cable spreading room and the 1F switchgear room would be limited to operator rounds and su rveillance activities.

Should any of these specific compensatory measures be discovered not in effect, the requirements of NMP-GM-031 (Reference 6) and associated instructions ensure appropriate plant risk mitigation is addressed. In this condition, compliance with NMP-GM-031 is considered compliance with the compensatory measures required by the proposed TS Required Actions.

E-10 Enclosure to NL-22-0316 Description and Assessment of the Proposed Change

Protected Equipment is controlled by governing SNC Procedure NMP-OS-010 (Reference 7). Systems or components that have been identified as essential to ensuring that safety functions or unit generation is maintained for given plant conditions, which includes systems supporting the primary system. Physical barriers or signage is used to alert personnel to maintain a safe distance from the Protected Equipment to prevent unintended consequences from operation, maintenance, or nearby activity.

Protected equipment has a physical barrier preventing access or work in the area and requires shift manager permission to enter the area. Operations personnel performing operator rounds are allowed to enter areas of protected equipment to ensure equipment conditions remain in the expected condition. In addition, operators monitor plant equipment to ensure no unauthorized work and periodically walk down postings and spot check behaviors and conditions to support effective equipment protection.

3.4 Maintenance Rule Control The PSW pumps are included under the HNP Maintenance Rule Program. The PSW pumps are monitored for unavailability as part of the Maintenance Rule performance monitoring. As part of compliance with 10 CFR 50.65, performance is monitored against licensee-established goals. If the performance of the PSW System does not meet the established goals, 10 CFR 50.65(a)(1) requires appropriate corrective action to be taken to restore the systems performance to an acceptable level.

Pump reliability is tracked by quarterly in-service testing (IST). If, during testing, pump parameters are outside of the established crit eria of the IST program, the IST program requires action to address the situation.

3.5 Evaluation of Risk Impacts The risks associated with a one-time extension of the HNP Unit 1 TS 3.7.2, Plant Service Water (PSW) System and Ultimate Heat Sink (UHS), Condition A Required Actions, to allow a one-time increase in the Completion Time from 30 days to 45 days have been evaluated by way of probabilistic risk assessment (PRA) models that meet all scope and quality requirements in NRC Regulatory Guide (RG) 1.200, Revision 2 (Reference 8).

This plant-specific risk assessment followed the guidance in RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3 (Reference 9), and RG 1.177, An Approach for Plant-Specific, Risk-Informed Decision Making: Technical Specifications, Revision 2 (Reference 1).

Attachment 4 of this license amendment request presents the evaluation of risk impacts due to the proposed amendment.

E-11 Enclosure to NL-22-0316 Description and Assessment of the Proposed Change

4. Regulatory Evaluation 4.1 Applicable Regulatory Requirements/Criteria The following regulatory requirements have been considered:

10 CFR 50.36:

10 CFR, Section 50.36, Technical specifications, in which the Commission established its regulatory requirements related to the contents of the TS: Specifically, 10 CFR 50.36(c)(2) states, in part, Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

The design of the PSW System satisfies 10 CFR 50.36, Technical Specifications, paragraph (c)(2)(ii), Criterion 3, which states the following:

(ii) A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The PSW System is described in the HNP Unit 1 UFSAR Section 10.7 (Reference 2).

The proposed amendment does not delete requirements associated with the PSW System and LCO 3.7.2 continues to maintain requirements associated with structures, systems, and components that are part of the primary success path and actuate to mitigate the related design basis accidents and transients.

The proposed amendment does not alter the remedial actions or shutdown requirements required by 10 CFR 50.36(c)(2)(i). Specifically, 10 CFR 50.36(c)(2)(i) states that:

Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condi tion for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

The proposed change meets this regulatory requi rement. When the limiting condition for operation (LCO 3.7.2) is not met, SNC will either shutdown the reactor (TS 3.7.2 Condition E) or follow remedial action permitted by the TS (TS 3.7.2 Condition A Required Actions) until the condition can be met. Thus, the proposed change does not affect compliance with this regulation.

E-12 Enclosure to NL-22-0316 Description and Assessment of the Proposed Change

Atomic Energy Commission (AEC) Design Criteria Following implementation of the proposed change, HNP Unit 1 will remain in compliance with applicable Atomic Energy Commission (AEC) design criteria as described in the HNP Unit 1 UFSAR (Reference 10).

4.2 Precedent The proposed amendment is similar to NRC-approved License Amendment 311 issued to SNC on September 24, 2021, for HNP Unit 1 (NRC Agencywide Documents Access and Management System (ADAMS) Accessi on No. ML21264A644). HNP Unit 1 Amendment 311 revised TS 3.7.2, Plant Service Water (PSW) System and Ultimate Heat Sink (UHS), Required Actions for Condition A, One PSW pump inoperable, to allow a one-time increase in the Completion Time from 30 days to 45 days during a specified time and only applicable to the HNP 1C PSW Pump with compensatory measures in place.

4.3 No Significant Hazards Consideration Analysis Pursuant to 10 CFR 50.90, Southern Nuclear Operating Company (SNC) hereby requests an amendment to Hatch Nuclear Plant (HNP) Unit 1 renewed facility operating license DPR-57. The proposed amendment would revise Condition A Required Actions, of Technical Specification (TS) 3.7.2, Plant Service Water (PSW) System and Ultimate Heat Sink (UHS), to extend the Completion Time from 30 days to 45 days for the 1A PSW Pump only. This proposed allowance would expire on May 19, 2022, at 1527 eastern daylight time (EDT) and be effective only while the compensatory measures described in Section 3.3 of this application are implemented.

SNC 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 involves a one-time extension to the Completion Time for TS 3.7.2 Condition A Required Actions to allow necessary time to restore the 1A PSW Pump to operable status. The proposed amendment does not affect accident initiators or precursors nor adversely alter the design assumptions, conditions, and configuration of the facility. The proposed amendment does not alter any plant equipment or operating practices with respect to such initiators or precursors in a manner that the probability of an accident is increased. The proposed amendment will not alter assumptions relative to the mitigation of an accident or transient event. Furthermore, the PSW System will remain capable of adequately responding to a design basis event or transient during the period of the extended Completion Time.

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

E-13 Enclosure to NL-22-0316 Description and Assessment of the Proposed Change

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

Response: No The proposed amendment does not introduce any new or unanalyzed modes of operation. The proposed changes do not involve a physical alteration to the plant (i.e., no new or different type of equipment will be installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made in the safety analysis.

Therefore, the proposed amendment will not create the possibility of a new or different accident from any accident previously evaluated.

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

Response: No The margin of safety is related to 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. The performance of these fission product barriers is not affected by the proposed amendment; therefore, the margins to the onsite and offsite radiological dose limits are not significantly reduced.

In addition, during the extended Completion Time, the PSW System will remain capable of providing the required cooling to systems responsible for mitigating the consequences of a design basis event such as a loss of coolant accident.

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

Based on the above, SNC 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.

4.4 Conclusions In conclusion, based on the considerations discussed herein, (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 Commissions 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.

E-14 Enclosure to NL-22-0316 Description and Assessment of the Proposed Change

5. Environmental Consideration SNC has evaluated the proposed amendment and has determined that the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types, or significant increase in the amounts, of any effluent that may be released off site, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the e ligibility criteria 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.

6. References

1. Regulatory Guide 1.177, An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications, Revision 2, January 2021

2. HNP-1-FSAR, Edwin I. Hatch Nuclear Plant Final Safety Analysis Report, Revision 38, September 2020, Section 10.7, Plant Service Water System

3. SNC Procedure Instruction NMP-OS-010-002, Hatch Protected Equipment Logs,

Version 12.1, Effective March 9, 2022

4. SNC Procedure 34AB-Y22-002-0, Naturally Occurring Phenomena, Version 20.1, Effective November 22, 2021

5. SNC Procedure 34AB-P41-001-1, Loss of Plant Service Water, Version 11.10, Effective January 1, 2022

6. SNC Procedure NMP-GM-031, On-Line Configuration Risk Management Program,

Version 9.0, Effective February 23, 2021

7. SNC Procedure NMP-OS-010, Protected Train/Division and Protected Equipment Program, Version 8.1, Effective March 23, 2020

8. Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2, March 2009

9. Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3, January 2018

10. HNP-1-FSAR, Edwin I. Hatch Nuclear Plant Final Safety Analysis Report, Revision 38, September 2020, Section F.5, Evaluation with Respect to 1971 General Design Criteria

E-15 Edwin I. Hatch Nuclear Plant Unit 1

Emergency License Amendment Request for Technical Specification 3.7.2 Regarding One-Time Extension of Completion Time for Plant Service Water (PSW) Pump Inoperable

Attachment 1

HNP Unit 1 Technical Specification Marked-up Page

PSW System and UHS 3.7.2 3.7 PLANT SYSTEMS

3.7.2 Plant Service Water (PSW) System and Ultimate Heat Sink (UHS)

LCO 3.7.2 Two PSW subsystems and UHS shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

A. One PSW pump inop erable. A.1 Restore PSW pump to 30 days OPERABLE status.

OR

NO TES----------------

1. Only applicable during 1C 1A PSW pump repair.

2. Only applicable until October 10, 2021May 19, 2022 at 16201527 EDT.

A.2.1 Establ ish compensatory 30 days measures as described in letter NL-21-0862 dated September 23, 2021, Enclosure 5NL-22-031 6 dated April 30, 2022, Enclosure Section 3.3.

AND A.2.2 Restore PSW pump to 45 days OPERABLE status.

B. One PSW turbine building B. 1 Restore PSW turbine 30 days isolation valve inoperable. building isolation valve to OPERABLE status.

C. One PSW pump in each C.1 Restore one PSW pump 7 days subsystem inoperable. to OPERABLE status.

D. One PSW turbine bui lding D.1 Restore one PSW turbine 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months isolation valve in each building isolation valve to subsystem inoperable. OPERABLE status.

(continued)

HATCH UNIT 1 3.7-3 Amendment No. 311 Edwin I. Hatch Nuclear Plant Unit 1

Emergency License Amendment Request for Technical Specification 3.7.2 Regarding One-Time Extension of Completion Time for Plant Service Water (PSW) Pump Inoperable

Attachment 2

HNP Unit 1 Revised Technical Specification Page

PSW System and UHS 3.7.2 3.7 PLANT SYSTEMS

3.7.2 Plant Service Water (PSW) System and Ultimate Heat Sink (UHS)

LCO 3.7.2 Two PSW subsystems and UHS shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

A. One PSW pump inop erable. A.1 Restore PSW pump to 30 days OPERABLE status.

OR

NO TES----------------

1. Only applicable during 1A PSW pump repair.

2. Only applicable until May 19, 2022 at 1527 EDT.

A.2.1 Establ ish compensatory 30 days measures as described in letter NL-22-031 6 dated April 30, 2022, Enclosure Section 3.3.

AND

A.2.2 Restore PSW pump to 45 days OPERABLE status.

B. One PSW turbine building B. 1 Restore PSW turbine 30 days isolation valve inoperable. building isolation valve to OPERABLE status.

C. One PSW pump in each C.1 Restore one PSW pump 7 days subsystem inoperable. to OPERABLE status.

D. One PSW turbine bui lding D.1 Restore one PSW turbine 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months isolation valve in each building isolation valve to subsystem inoperable. OPERABLE status.

(continued)

HATCH UNIT 1 3.7-3 Amendment No.

Edwin I. Hatch Nuclear Plant Unit 1

Emergency License Amendment Request for Technical Specification 3.7.2 Regarding One-Time Extension of Completion Time for Plant Service Water (PSW) Pump Inoperable

Attachment 3

HNP Unit 1 Technical Specification Bases Marked-up Page (information only)

PSW System and UHS B 3.7.2

BASES

APPLICABILITY The LCO for the PSW System and UHS is not applicable in MODES (continued) 4 and 5, and defueled. However, portions of the PSW System and UHS may be required to perform necessary support functions for OPERABILITY of the supported systems. Thus, the LCOs of the individual systems, which require portions of the PSW System and the UHS to be functional to support individual system OPERABILITY, will govern PSW System and UHS requirements during operation in MODES 4 and 5 and defueled.

ACTIONS A.1

With one PSW pump inoperable, the inoperable pump must be restored to OPERABLE status within 30 days. With the unit in this condition, the remaining OPERABLE PSW pumps (even allowing for an additional sin gle failure) are adequate to perform the PSW heat removal function; however, the overall reliability is reduced. The 30 day Completion Time is based on the remaining PSW heat removal capability to accommodate additional single failures, and the low probability of an event occurring during this time period.

A.2.1 and A.2.2

The Completion Time to restore one PSW pump to OPERABLE status to facilitate the 1C 1A PSW pump repair may be extended to 45 days total, provided action is taken within 30 days to establish compensatory and risk management controls.

The A.2.1 and A.2.2 Required Actions are modified by two Notes.

Note 1 ensures that the A.2.1 and A.2.2 Required Actions are only applied during the 1C 1A PSW pump repair. Note 2 limits the time period the A.2.1 and A.2.2 Required Actions may be used.

The extended Completion Time is subject to additional compensatory controls specified in SNC letter NL 0862NL 0316, dated April 30, 2022 September 23, 2021, that consist of controls that must be established and maintained during the extended C ompletion Time.

These controls are based on procedural protection, operation of redundant functions, and recommended actions based on risk insights.

If Required Action A.2.1 is met, the allow ed time to restore the PSW pump to OPERABLE status can be extended to 45 days from entry into Condition A. With the unit in this condition, the remaining OPERABLE PSW pumps (even allowing for an additional single (continued)

HATCH UNIT 1 B 3.7-9 REVISION 117 PSW System and UHS B 3.7.2

BASES

ACTIONS A.2.1 and A.2.2(continued)

failure) are adequate to perform the PSW heat removal function; however, the overall reliability is reduced. The 45-day Completion Time is based on the remaining PSW heat removal capability to accommodate additional single failures, the low probability of an event occurring during this time period, and the established compensatory measures of SNC letter NL-21-086222-0316.

B.1

With one PSW turbine building isolation valve inoperable, the inoperable valve must be restored to OPERABLE status wit hin 30 days. With the unit in this conditi on, the remaining OPERABLE PSW turbine building isolation valve in the subsystem is adequate to isolate the non-essential loads, and, even allowing for an additional single failure, the other PSW subsystem is adequate to perform the PSW heat removal function; however, the overall reliability is reduced.

The 30 day Completion Time is based on the remaining PSW heat removal capability to accommodate additional single failures, and the low probability of an event occurring during this time period.

(continued)

HATCH UNIT 1 B 3.7-9a REVISION 117 Edwin I. Hatch Nuclear Plant Unit 1

Emergency License Amendment Request for Technical Specification 3.7.2 Regarding One-Time Extension of Completion Time for Plant Service Water (PSW) Pump Inoperable

Attachment 4

Evaluation of Risk Impact and Compensatory Measures

NL-22-0316, Attachment 4 Evaluation of Risk Impact and Compensatory Measures

1.0 INTRODUCTION

1.1 PURPOSE The purpose of this analysis is to assess the acceptability, from a risk perspective, o f a change to extend the Hatch completion time (CT) for Unit 1 Tech Spec Condition 3.7.2 Required Action A.1 from 30 days to 45 days in order to allow for repair of the 1A Plant Service Water (PSW) Pump.

These proposed changes are requested to be effective only during a one-time extension.

1.2 BACKGROUND

1.2.1 Technical Specification Changes Since the mid-1980s, the NRC has been reviewing and granting improvements to TS that are based, at least in part, on probabilistic risk assessment (PRA) insights. In its final policy statement on TS improvements of July 22, 19 93, the NRC stated that it...

... expects that licensees, in preparing their Technical Specification related submittals, will utilize any plant-specific PSA or risk survey and any available literature on risk insights and PSAs... Similarly, the NRC s taff will also employ risk insights and PSAs in evaluating Technical Specifications related submittals.

Further, as a part of the Commissions ongoing program of improving Technical Specifications, it will continue to consider methods to make better use of risk and reliability information for defining future generic Technical Specification requirements.

The NRC reiterated this point when it issued the revision to 10 CFR 50.36, Technical Specifications, in July 1995. In August 1995, the NRC adopted a final policy statement on the use of PRA methods in nuclear regulatory activities that encouraged greater use of PRA to improve safety decision-making and regulatory efficiency. The PRA policy statement included the following points:

1. The use of PRA technology should be increased in all regulatory matters to the extent supported by the state of the art in PRA methods and data and in a manner that complements the NRCs deterministic approach and supports the NRCs traditional defense-in-depth philosophy.

2. PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, and importance measures) should be used in regulatory matters, where practical within the bounds of the state of the art, to reduce unnecessary conservatism associated with current r egulatory requirements.

3. PRA evaluations in support of regulatory decisions should be as realistic as practicable and appropriate supporting data should be publicly available for review.

Attachment 4, Page 1 NL-22-0316, Attachment 4 Evaluation of Risk Impact and Compensatory Measures

4. The Commission's safety goals and subsidiary numerical objectives are to be used with consideration of uncertainties in making regulatory judgments The movement of the NRC to more risk-informed regulation has led to the NRC identifying Regulatory Guides and associated processes by which licensees can submit changes to the plant design basis including Technical Specif ications. These guides are discussed in the following section.

1.3 REGULATORY GUIDES Three Regulatory Guides provide primary inputs to the evaluation of a Technical Specification change. Their relevance is discussed in this section.

1.3.1 Regulatory Guide 1.200, Revision 2 Regulatory Guide 1.200, Revision 2 [Ref. 1] describes an acceptable approach for determining whether the quality of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision-making for light-water reactors. This guidance is intended to be consistent with the NRCs PRA Policy Statement and more detailed guidance in Regulatory Guide 1.174.

It is noted that RG 1.200, Revision 2 endorses Addendum A of the ASME/ANS PRA Standard

[Ref. 4] as clarified in Appendix A of RG 1.200, Revision 2.

1.3.2 Regulatory Guide 1.174, Revision 3 Regulatory Guide 1.174 [Ref. 2] specifies an approach and acceptance guidelines for use of PRA in risk informed activities. RG 1.174 outlines PRA related acceptance guidelines for use of PRA metrics of Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) for the evaluation of permanent TS changes. The guidelines given in RG 1.174 for determining what constitutes an acceptable permanent change specify that the CDF and the LERF associated with the change should be less than specified values, which are dependent on the baseline CDF and LERF, respectively.

RG 1.174 also specifies guidelines for consideration of external events. External events can be evaluated in either a qualitative or quantitative manner.

Since this LAR is for a one-time TS change, the CDF and the LERF of RG 1.174 do not specifically apply.

1.3.3 Regulatory Guide 1.177 Revision 2 Regulatory Guide 1.177 [Ref. 3] specifies a risk-informed approach and acceptance guidelines for the evaluation of plant technical specification changes. RG 1.177 identifies a three -tiered approach for the evaluation of the risk associated with a proposed TS change as identified below:

Tier 1 is an evaluation of the plant-specific risk associated with the proposed TS change, as shown by the change in core damage frequency (CDF) and incremental conditional core damage probability (ICCDP). W here

Attachment 4, Page 2 NL-22-0316, Attachment 4 Evaluation of Risk Impact and Compensatory Measures

applicable, containment performance should be evaluated on the basis of an analysis of large early release freq uency (LERF) and incremental conditional large early release probability (ICLERP). The acceptance guidelines given in RG 1.177 for determining an acceptable permanent TS change are that the ICCDP and the ICLERP associated with the change should be less than 1E-06 and 1E-07, respectively. RG 1.177 also addresses risk metric requirements for one-time TS changes, as outlined in Section 1.3.4 (Acceptance Guidelines) of this risk assessment.

Tier 2 identifies and evaluates, with respect to defense-in-depth, any potential risk-significant plant equipment outage configurations associated with the proposed change. The licensee should provide reasonable assurance that risk-significant plant equipment outage configurations will not occur when equipment associated with the proposed TS change is out-of-service.

Tier 3 provides for the establishment of an overall configuration risk management program (CRMP) and confirmation that its insights are incorporated into the decision-making process before taking equipment out-of-service prior to or during the CT. Compared with Tier 2, Tier 3 provides additional coverage based on any additional risk significant configurations that may be encountered during maintenance scheduling over extended periods of plant operation. Tier 3 guidance can be satisfied by the Maintenance Rule (10 CFR 50.65(a)(4)), which requires a licensee to assess and manage the increase in risk that may result from activities such as surveillance, testing, and corrective and preventive maintenance.

This risk analysis supports the three tiers of RG 1.177, specifically the comparison of the results with the acceptance guidelines for ICCDP and ICLERP associated with changing a Technical Specification Completion Time, the assessment of risk-significant combinations, and the use of the Configuration Risk Management Program.

1.3.4 Acceptance Guidelines Risk significance in a LAR is determined by comparison of changes in Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) and values of Incremental Conditional Core Damage Probability (ICCDP) and Incremental Conditional Large Early Release Probability (ICLERP) produced by a permanent change to either the plant design basis or Technical Specifications to the guidelines given in Regulatory Guide 1.174 and Regulatory Guide 1.177.

Regulatory Guide 1.174 specifies the acceptable changes in CDF and LERF for permanent changes. Regulatory Guide 1.177 specifies the acceptable ICCDP and ICLERP for temporary changes, usually associated with changing CT.

Regulatory Guide 1.177 directly addresses the risk metric requirements for one-time TS changes, as reproduced below:

For one-time only changes to TS CTs, the frequency of entry into the CT may be known, and the configuration of the plant SSCs may be established. Further, there is no permanent change to the plant CDF or LERF, and hence the risk

Attachment 4, Page 3 NL-22-0316, Attachment 4 Evaluation of Risk Impact and Compensatory Measures

guidelines of Regulatory Guide 1.174 cannot be applied directly. The following TS acceptance guidelines specific to one -time only CT changes are provided for evaluating the risk associated with the revised CT:

1. The licensee has demonstrated that implementation of the one-time only TS CT change impact on plant risk from implementing the one-time only TS CT change is acceptable (Tier 1):

An ICCDP of less than 1.0x10-6 and an ICLERP of less than 1.0x10 -7, or

An ICCDP of less than 1.0x10 -5 and an ICLERP of less than 1.0x10 -6 with effective compensatory measures implemented to reduce the sources of increased risk.

2. The licensee has demonstrated that there are appropriate restrictions on dominant risk-significant configurations associated with the change (Tier 2).

3. The licensee has implemented a risk-informed plant configuration control program. The licensee has implemented procedures to utilize, maintain, and control such a program (Tier 3).

Based on the available quantitative guidelines for other risk-informed applications, it is judged that the quantitative criteria shown in Table 1-1 represent a reasonable set of acceptance guidelines. For the purposes of this evaluation, these guidelines demonstrate that the risk impacts are acceptably low. This, combined with effective compensatory measures to maintain lower risk, will ensure that the TS change meets the intent of small risk increases consistent with the Commission's Safety Goal Policy Statement.

Table 1-1 PROPOSED RISK ACCEPT ANCE GUIDELINES

RISK ACCEPT ANCE GUIDELINE BASIS

ICCDP < 1E-6, or ICCDP is an appropriate metric for ICCDP < 1E-5 with effective compensatory assessing risk impacts of out of service measures implemented to reduce the sources equipment per RG 1.177. This guideline is of increased risk specified in Section 2.4 of RG 1.177.

ICLERP < 1E-7, or ICLERP is an appropriate metric for ICLERP < 1E-6 with effective compensatory assessing risk impacts of out of service measures implemented to reduce the sources equipment per RG 1.177. This guideline is of increased risk specified in Section 2.4 of RG 1.177.

Attachment 4, Page 4 NL-22-0316, Attachment 4 Evaluation of Risk Impact and Compensatory Measures

1.4 SCOPE This section addresses the requirements of RG 1.200, Revision 2 which directs the licensee to define the treatment of the scope of risk contributors (i.e., internal initiating events, external initiating events, and modes of power operation at the time of the initiator). Discussion of these risk contributors are as follows:

Internal Events (IE) - The Hatch PRA model used for this analysis includes a full range of internal initiating events for at-power configurations. Loss of PSW is a modeled special initiating event and logic for this is in the model.

The IE model is further discussed in Section 1.5.

Internal Flooding (IF) - The Hatch PRA model used for this analysis includes flooding scenarios. The IF model is further discussed in Section 1.5.

Low Power Operation - The intent is for the unit to remain at power during the completion time. PSW provides motor cooling for RHR Service Water pumps during shutdown. Since RHRSW is used for shutdown cooling, and RHRSW is cooled by PSW, there is some risk involved with going into lower modes; however, that is not quantified or discussed any further in this assessment. As described below, since the shutdown success criteria only requires one pump, and three are still operable, this is a small increase only.

Shutdown / Refueling - Hatch does not have a shutdown PRA model, but instead relies upon NUMARC 91-06 deterministic methodology to assess defense-in-depth of key safety functions. PSW is not measured directly but is considered a support system for the key safety functions.

Internal Fires - The Hatch PRA model used for this analysis contains an as-built, as-operated Fire PRA model. (Note that Internal Fires are often included in the list of external events evaluated by a PRA; that is the case for Hatch as well.) The Fire PRA model is discussed further in Section 1.5.

Seismic - The Hatch PRA model used for this analysis includes a Seismic PRA. The SPRA is further discussed in Section 1.5.

Other External Events - Other external event risks (including external flooding and high winds) were assessed in the Hatch Other External Events Screening calculation [Ref. 6]. The Other External Events are further discussed in Section 2.3.

1.5 HATCH PRA MODELS This section addresses the requirements of Section 3.1 of RG 1.200, Revision 2 [Ref. 1] which directs the licensee to identify the portions of the PRA used in the analysis.

The PRA analysis uses Revision 8A Phoenix O ne-Top Multi-Hazard Model contained in SNC calculation RIE-PHOENIX-U01 [Ref. 5]. This model has the required quantification and support files set up to calculate either zero-maintenance or average maintenance risks. To clarify, it was used to generate average maintenance risk except where adjustments to components are

Attachment 4, Page 5 NL-22-0316, Attachment 4 Evaluation of Risk Impact and Compensatory Measures

specifically documented. It also implements several model enhancements identified (discussed in Risk Assessment Details) during PHOENIX development and therefore represents the most accurate model of record available. As described in RG 1.177, subsequent issues identified with the model would most likely impact the base and configuration specific models equally, therefore the delta risk calculations for a one-time TS change should not be impacted. If a permanent change were requested, model issues could impact the overall CDF and LERF and would need to be addressed further. Even so, uncertainty associated with some items that are not currently modeled are addressed (Open Phase, Breaker Coordination).

The Revision 8A Phoenix OTMHM model of record contains internal events, internal flooding, internal fire, and seismic hazards. All other hazards screened out as being very low risk. There is an open item related to the TSTF-505 audit for the tornado hazard, which is further addressed in Section 2.3. The model can be evaluated one hazard at a time or with all hazards activated.

Each hazard model has been peer reviewed against the ASME peer review standard, and all the F&Os have been addressed. There are two open findings related to internal flooding documentation that do not impact the outcome of this assessment.

F&O 1-9: This is related to SRs AS-B3 and AS-C2. The F&O is related to missing discussion on the phenomenological conditions expected for each accident sequence related to SBO with usage of fire water. Based on the Hatch Equipment Qualification Program, equipment located in potentially harsh environment conditions, including inside containment, are expected to perform its safety function when exposed to normal, abnormal, and accident environment. For all other areas, the models do not credit use of equipment in the area of events that cause adverse environmental events, such as ISLOCA events and steam line breaks outside containment. The Internal flooding analysis evaluates the susceptibility of components to spray and flooding separately. A discussion on environmental considerations for the SBO sequence where fire water is used was added to the documentation. This finding is a documentation issue; there is no impact on the unavailability of 1A PSW pump being out of service.

F&O 6-8: This is related to SR HR-G6. The finding was related to the Hatch Human Reliability Analysis document where the consistency check did not include comparison of HEPs in regards to scenarios context, plant history, procedures, operational practices, and experience. The internal events HRA documentation was revised to incorporate a better consistency analysis. A discussion on requirements from NUREG-1792 a nd feasibility requirements from NUREG-1921 to be used for internal events HFEs were added to the HRA notebook. HFEs and their HEP were reviewed relative to each other to check their reasonableness given the scenario context including plant procedures, plant history, operational practices and experiences and documented in the HRA notebook. Thus the documentation associated with this issue has been revised. There is no impact on the unavailability of 1A PSW pump being out of service.

A review of the quantification and uncertainty notebooks for each hazard model did not find any assumption or uncertainty that would impact the results of this evaluation.

PRA Maintenance and Updates

SNC Risk Informed Engineering (RIE) developed a comprehensive PRA model and application maintenance process in response to internal and external assessments and issuance of industry

Attachment 4, Page 6 NL-22-0316, Attachment 4 Evaluation of Risk Impact and Compensatory Measures

configuration management guidance documents. This process ensures that the applicable PRA models remain as accurate reflections of the as-built and as-operated units. This process delineates the responsibilities and guidelines fo r updating the PRA models at all operating SNC nuclear generation sites. It defines the process for implementing PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, industry operational experience), and for controlling the model and associated computer files. Components of this process include:

Design change impact reviews are performed by RIE prior to implementation.

Procedures that can affect PRA modeling or assumptions are reviewed by RIE prior to issue.

Licensing document changes are reviewed by RIE prior to issue.

RIE personnel participate in the Maintenance Rule expert panel, Surveillance Frequency Control Program and 10CFR50.69 Independent Decision-Making Panels.

SNC Procedure RIE-001 requires that potential impacts to the PRA models be identified and entered in the PRA Model Change log. Each entry in the change log requires an evaluation of the impact of the individual change, as well as an evaluation of the cumulative impact for unincorporated changes. This results in a continuous change tracking process so that the difference between the models and the plant can be quickly determined and evaluated.

In addition to these activities, SNC risk m anagement procedures provide guidance for PRA documentation quality and maintenance activities. This guidance includes:

Documentation of the PRA model, PRA products, and bases documents.

Requirement to evaluate model changes against the ASME standard definitions of Upgrade and Model Maintenance. Requirement to conduct focused peer review for any changes classified as an Upgrade.

The approach for controlling electronic storage of Risk Management (RM) products including PRA update information, PRA models, and PRA applications.

Guidance for use of quantitative and qualitative risk models in support of the On-Line Work Control Process Program for risk evaluations for maintenance tasks (corrective maintenance, preventive maintenance, minor maintenance, surveillance tests and modifications) on systems, structures, and components (SSCs) within the scope of the Maintenance Rule (10 CFR 50.65 (a)(4)).

In accordance with this guidance, regularly scheduled PRA model updates nominally occur on an approximate two refueling outage cycle; however, longer intervals may be justified if it can be shown that the PRA continues to adequately represent the as-built, as-operated plant.

Attachment 4, Page 7 NL-22-0316, Attachment 4 Evaluation of Risk Impact and Compensatory Measures

2.0 RISK ANALYSIS This section evaluates the plant-specific risk associated with the proposed TS change, based on the risk metrics of CDF, ICCDP, LERF, and ICLERP.

2.1 ASSESSMENT OVERVIEW AND ASSUMPTIONS 2.1.1 Overview This analysis is performed for unavailability of 1A PSW Pump. The PRA analysis involves identifying the system and components or maintenance activities modeled in the PRA which are most appropriate for use in representing the extended CT configurations and comparing the results to the baseline. Table 2.1-1 lists the base risk metrics for the Full Power Internal Events (FPIE) PRA, internal flooding PRA, Seismic PRA (SPRA) and the Fire PRA (FPRA).

Table 2-1 HATCH CDF AND LERF BASE RISK METRICS

Hazard(s) Risk (1/yr.)

OTMHM CDF 4.8E-05

OTMHM LERF 3.0E-06

The general configuration for the extended CT is Hatch U1 at-power with the 1A PSW Pump out of service. Additional adjustments are described for each case ran. The risk impact is for Unit

1. The planned maintenance is expected to focus on repairing the pump within the requested extended CT. Concurrent maintenance work will be carefully managed during the extended CT, using the Configuration Risk Management Program and compensatory measures.

The PRA model was quantified using the base average test and maintenance PRA model with the 1A PSW Pump basic events set to TRUE. This included other currently out of service plant equipment. The average test and maintenance model represent baseline assumed maintenance frequencies for all components except for Technical Specification violations that ar e normally excluded in the disallowed maintenance (mutually exclusive) logic in the base PRA model. As a conservative measure, maintenance events for equipment that is protected per site processes during the extended completion time was left at their nominal values.

Table 2-2 EXTENDED CT CONFIGURATION OUT OF SERVICE REPRESENTATION Component Basic Event Description 1R22S001/CB5 CBFO1R22S001_5 1R22S001 Normal Supply Breaker 1X43C002B CC-DFPR-6, CC-DFPS-5, Diesel Fire Pump B HVXC1X43F351K 1P41F017A HVXC1P41F017A PSW to Drywell Cooler B007A

Attachment 4, Page 8 NL-22-0316, Attachment 4 Evaluation of Risk Impact and Compensatory Measures

Component Basic Event Description 1P41F053A HVXC1P41F053A PSW to Drywell Cooler B007A 1P41F397B HVXC1P41F397B PSW to heat exchanger B 1

1P41F398B HVXC1P41F398B PSW to heat exchanger B 1 1P41F399B HVXC1P41F399B PSW to heat exchanger B 1 1P42F001B HVXC1P42F001B RBCCW to heat exchanger B 1 1P42F002B HVXC1P42F002B RBCCW to heat exchanger B 1 CC-PS-4 1A PSW Pump fails to run for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months CC-SW-1 1A PSW Pump fails to start MNUNPS_TRNA 1A PSW Pump in maintenance CC-PS-4________I 1A PSW Pump fails during the year as part of the Loss of PSW modeled initiating event, set to False for this Basic evaluation events directly associated CC-SW-34 PSW 1A outlet check valve with 1A PSW Pump fails to close following PSW 1A fails to run (PSW 1C flow diversion), set to False for this evaluation HVXC1P41F009A cooling water to pump motor inlet HVXC1P41F020A cooling water to pump motor outlet HVXC1P41F302A pump discharge manual isolation valve 1X41C009B CC-INT-5, FNOS1X41C009B intake structure vent fan, located in roof plug removed for pump access

1) Leaking RBBCW Heat Exchanger Tubes

The PRA model success criteria for PSW pumps is that 1 of 4 pumps is needed to respond to transient events, and 3 of 4 pumps are needed to prevent the Loss of PSW initiating event and keep the unit on-line.

To produce a conservative result, the maintenance terms for components that are protected per the protected train procedure NMP-OS-010 are left at their nominal values. Since planned maintenance on these components will not be performed while 1A PSW Pump is out of service, this result is conservative.

Attachment 4, Page 9 NL-22-0316, Attachment 4 Evaluation of Risk Impact and Compensatory Measures

1P41C001C - Control switch, 4160v breaker and pump.

A flag file was used to change the event probabilities of the impacted events. The base model OTMHM flag file was used as a starting point and the above events were added to it.

Common Cause event adjustments due to compo nent down for maintenance RG 1.177 contains specific directions on adjusting the CCF events in a model if the component is unavailable due to schedule maintenance and not failed. The CCF contributions involving the component should be modified to remove the component and to only include failures of the remaining components. However, CCF was not adjusted and CCF events involving the 1A PSW Pump are contained in cutsets. This is conservative because in reviewing the results, there are common cause groups in cutsets that are not possible with the 1A PSW Pump outage. When reviewing the importance measures, the determination of compensatory measures was not impacted by these common cause groups.

The three other pumps are in reliable performance conditions based on plant health summary report. This issue would be observed during post-maintenance testing the next time the pump and/or motor is replaced.

2.1.2 Quantification Truncation To generate both the base and CT case risk, each hazard was quantified at the truncation levels below to ensure that the basic events for the 1A PSW pump were present.

Internal Events CDF - 1E-12 Internal Events LERF - 1E-13 Internal Flooding CDF - 1E-12 Internal Flooding LERF - 1E-14 Internal Fire CDF - 1E-11 Internal Fire LERF - 1E-11 Seismic CDF - 1E-11 Seismic LERF - 1E-11

2.1.3 Calculation Approach Evaluation Assumptions A) No quantitative credit is taken for the protection of equipment per site procedure NMP-OS-010. Since routine work on a number of components is prohibited per this process, the maintenance events for these could be set to zero in the assessment. This was not done and thus the results are conservative.

B) Existing components already out of service are not included in the base risk calculations.

This results in a lower base risk and therefore a higher delta risk; thus, the risk calculations are conservative.

C) Credit for FLEX equipment is turned off in both the base case and 1A PSW OOS case models.

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Risk Assessment Details The base model used for this assessment is the Revision 8A Phoenix one-top model of record documented in calculation H-RIE-PHOENIX-U01, Revision 4. This model has the required quantification and support files set up to calculate either zero-maintenance or average maintenance risks. It contains flag events that allow NFPA-805 credited modifications to be turned on or off. This risk assessment takes no credit for NFPA-805 modifications yet to be installed.

Updates that have taken place since the 10 CFR 50.69 and NFPA-805 approvals were documented in the Hatch 1B Diesel Generator Liner Replacement One-Time Technical Specification Extension LAR and RAIs. The NRC sta ff found the changes to the PRA model to be adequately described and justified to support t he risk analysis for the EDG one-time AOT extension. The second set of updates to the Hatch PRA Model was in preparation of the Hatch Risk Informed Completion Time (RICT) LAR. All the updates to the PRA for the Hatch RICT LAR were determined to be maintenance updates and not PRA upgrades. After the Hatch RICT LAR Audit, an investigation was made into why there was a negative delta risk value in the Hatch DG Liner LAR for the quantification of selected hazards (e.g., seismic). As discussed below, the changes were made to the PRA model as a result of the investigation. This change and the change to the flag file to take credit for NFPA 805 modification installed during the previous outage are the only changes made to the Hatch PRA model following the Hatch RICT LAR Audit.

These changes are determined to be PRA maintenance and not PRA upgrades.

Currently FLEX equipment and related actions are only modeled in the Seismic PRA model.

Review of the fault tree used to calculate the Base and scenario specific CDF and LERF for all hazards confirmed that the model flag FL-FLEX is set to credit FLEX for Seismic PRA only. When the credit for FLEX is removed in the Seismic PRA by setting the basic event OPHEELAP-COG-S (operator fails to implement the ELAP procedure) to 1.0 and setting the seismic versions of this basic event (SRX*_ OPHEELAP-COG-S) to true, the resulting calculated delta risk results show an increase of 7E-8.

Revision 8A model flag file was adjusted to ac count for the NFPA 805 modification installed during the previous outage by setting the following modification flag events to false. This was done for the base case and the 1A PSW Pump outage case. The credit for these modifications decreased base case Fire CDF by 1.6E-5 and Fire LERF by 1.4E-6.

The recovery rule file applies a recovery fault tree to the cutsets and joint human failure probabilities based on the base case human reliability analysis dependency analysis. The recovery fault tree uses the same database as the fault tree used in quantification. The recovery fault tree is used for Consequential LOSP scenarios. The recovery fault tree and the main fault tree use the same database; no changes were made to the recovery fault tree structure or logic.

Failure of the 1A PSW Pump has no impact on the existing recovery rules. Revision 8A model recovery file was adjusted (Table 2) to add changes that came from the investigation into Hatch RICT LAR Audit Question 12. These changes were made in both the base case and the 1A PSW Pump outage case.

Table 2-3

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Recovery Rule File Changes from Hatch RICT LAR Audit Q12 Investigation Item Recovery Rule Change Compression of events in the recovery rule Added at the beginning file before applying recovery to cutsets. **SET EVENTS UNKNOWN**

These events are PROB in the flag file.

The location in the recovery rule file on Moved after the change from above setting sequence flags to TRUE and **SET EVENTS TRUE**

creating non-minimal - erroneous results - SEQ*

cutsets and causing results to differ FLG_OSPREC between the base and DG OOS cutsets.

- SUBSUME**

- COMPRESS** 4

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The location of the application-specific rule Moved after the change from above file and associated comments for removing ; Remove illegitimate cutsets. These cutsets and applying recovery to cutsets relevant have OSP recovery successful on the success to offsite power recovery. branch of the sequence but no credit for OSP recovery on the failure branch. IE and FLD only.

- RECOVERY** DEL 0

%IE* SEQ_LOSP-2 FLG_OSPREC OPHE*

%IE* SEQ_LOSP-5 FLG_OSPREC OPHE*

%FLD* SEQ_LOSP-2 FLG_OSPREC OPHE*

%FLD* SEQ_LOSP-5 FLG_OSPREC OPHE*

Added a Subsume/Compress 4 command after the recovery tree is called.

Moved after Seismic LERF delete commands

Apply OPHE-REC-PSW-F

- CHANGEEVENTS** +OPHE-REC-PSW-F

%HF_1101J* SEQ_SBO_28 -OPHE* -CC-SW-1

- CHANGEEVENTS** +OPHE-REC-PSW-F

%HF_1101J* SEQ_SBO_5 -OPHE* -CC-SW-1

- SET EVENT PROBS**

OPHE-REC-PSW-F.05

Moved before HRA Dependency Rules

These cutsets conservatively applied a 5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months OSP recovery when 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months is warranted. IE and FLD only

- MAX RECOVERIES** 1

- RECOVERY** NR-OSP-5to10H 0.523

%IE* SEQ_LOSP-9 CONSEQ-NR1-5H

%IE* SEQ_GT_10 CONSEQ-NR1-5H

%IE* SEQ_GT_3 CONSEQ-NR1-5H

%IE* SEQ_LOSP-13 CONSEQ-NR1-5H

%IE* SEQ_LOSP-16 CONSEQ-NR1-5H

%IE* SEQ_GT_16 CONSEQ-NR1-5H

%IE* SEQ_LOSP-22 CONSEQ-NR1-5H

%IE* SEQ_LOSP-19 CONSEQ-NR1-5H

%IE* SEQ_GT_22 CONSEQ-NR1-5H

%IE* SEQ_GT_19 CONSEQ-NR1-5H

%IE* SEQ_LOSP-9 ZNR-OSP-5H

%IE* SEQ_GT_10 ZNR-OSP-5H

%IE* SEQ_GT_3 ZNR-OSP-5H

%IE* SEQ_LOSP-13 ZNR-OSP-5H

%IE* SEQ_LOSP-16 ZNR-OSP-5H

%IE* SEQ_GT_16 ZNR-OSP-5H

%IE* SEQ_LOSP-22 ZNR-OSP-5H

%IE* SEQ_LOSP-19 ZNR-OSP-5H

%IE* SEQ_GT_22 ZNR-OSP-5H

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Item Recovery Rule Change

%IE* SEQ_GT_19 ZNR-OSP-5H (continued) **MAX RECOVERIES** 1

- RECOVERY** NR-OSP-5to10H 0.523

%FLD* SEQ_LOSP-9 CONSEQ-NR1-5H

%FLD* SEQ_GT_10 CONSEQ-NR1-5H

%FLD* SEQ_GT_3 CONSEQ-NR1-5H

%FLD* SEQ_LOSP-13 CONSEQ-NR1-5H

%FLD* SEQ_LOSP-16 CONSEQ-NR1-5H

%FLD* SEQ_GT_16 CONSEQ-NR1-5H

%FLD* SEQ_LOSP-22 CONSEQ-NR1-5H

%FLD* SEQ_LOSP-19 CONSEQ-NR1-5H

%FLD* SEQ_GT_22 CONSEQ-NR1-5H

%FLD* SEQ_GT_19 CONSEQ-NR1-5H

%FLD* SEQ_LOSP-9 ZNR-OSP-5H

%FLD* SEQ_GT_10 ZNR-OSP-5H

%FLD* SEQ_GT_3 ZNR-OSP-5H

%FLD* SEQ_LOSP-13 ZNR-OSP-5H

%FLD* SEQ_LOSP-16 ZNR-OSP-5H

%FLD* SEQ_GT_16 ZNR-OSP-5H

%FLD* SEQ_LOSP-22 ZNR-OSP-5H

%FLD* SEQ_LOSP-19 ZNR-OSP-5H

%FLD* SEQ_GT_22 ZNR-OSP-5H

%FLD* SEQ_GT_19 ZNR-OSP-5H

The Revision 8A Phoenix Risk Model of record (PRM) contains internal events, internal flooding, internal fire, and seismic hazards. All other hazards have been screened out as being very low risk. The model can be evaluated one hazard at a time or with all hazards activated.

PRAQuant 5.2 was used for quantification. FTREX 2.0.0.1 64-bit wrapper was used as the quantification engine. The cutsets were calculated without recoveries applied and then, due to the complexity of the recovery rules and the limitations of the 32-bit version of QRecover provided with PRAQuant., QRecover version 10 wa s used externally to apply recovery rules.

The base risk used for this ev aluation is taken directly from the PRM Revision 8A calculation.

Existing out of service components were not included in the base case quantification. The components currently out of service in addition to 1A PSW Pump were included only in the pump failed cases. This adds conservatism to the evaluation since the base case is lower and the delta risk is higher than if the out of service components were included in the base case.

The proposed technical specification change involves unavailability of the 1A PSW Pump. The revised CDF and LERF values for the CT configurations are obtained by re-quantifying the base PRA model with the identified events se t, as shown below, in a flag file.

The evaluation of ICCDP and ICLERP for this condition is determined as shown below:

The ICCDP associated with 1A PSW Pump OOS for a new CT is given by:

ICCDP1A= (CDF1A - CDFBASE) x CTNEW

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where CDF1A = the annual average CDF calculated with 1A PSW Pump OOS (and other currently OOS equipment)

CDFBASE = baseline annual average CDF with average unavail ability for all equipment. This is the CDF result of the baseline PRA (all quantified hazards). Currently OOS equipment was not included in the base case values.

CTNEW = the new extended CT (in units of years)

Note: ICCDP is a dimensionless probability.

Risk significance relative to ICLERP is determined using equations of the same form as noted above for ICCDP.

Since this evaluation is for a one-time TS CT allowance, the ICCDP and ICLERP are the only meaningful metrics as there is no permanent ch ange in plant risk after this one-time CT extension.

The guidance provided in Regulatory Guide 1.200, Revision 2, (Section C.4.2) requires the following items be addressed in documentation submitted to the NRC to demonstrate the technical adequacy of PRA models utilized for the application:

Identification of permanent plant changes (such as design or operational practices) that have an impact on the PRA but have not been incorporated in the PRA.

The parts of the PRA used to produce the results are performed consistently with the version of the PRA Standard endorsed by RG 1.200.

A summary of the risk assessment methodology used to assess the risk of the application, including how t he PRA model was modified to appropriately model the risk impact of the application.

Identifications of key assumptions and approximations in the PRA relevant to the results used in the decision-making process.

A discussion of the resolution of peer review or self-assessment findings and observations that are applicable to the parts of the PRA required for the application.

Identification of parts of the PRA used in the analysis that were assessed to have capability categories less than that required for the application.

Impact of potentially degraded performance of the in-service PSW pumps The 1A PSW Pump was taken out of service for scheduled motor replacement. Vibration data collected for the 1A PSW Pump after motor replacement was unacceptable. On April 26, 2022, following the torque of the motor, discharge head and pump column assembly, the discharge check valve was being bolted to the discharge head flange and a gap was encountered outside of desired parallel conditions.

With this being a maintenance issue in aligning the pump to the discharge check valve while trying to restore the pump to service, there is no impact of potential degraded performance for the other three PSW pumps from this condition. The three other pumps are in reliable

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performance conditions based on plant health summary report and reliability for those pumps has not been adjusted.

2.2 OTMHM Quantification The relevant inputs from the PRA models to Equation 2-1 (and the equivalent for LERF) are shown in Table 2.2-1 below.

Table 2-4 OTMHM Risk Assessment Parameters and Results for Unit 1 Integrated Value for Hazard Base risk (/yr.) New risk (/yr.) Delta Risk (/yr.) 15-day extension (45-day total CT)

Internal Events CDF 4.1E-06 6.6E-06 2.6E-06 Internal Flooding CDF 2.9E-07 3.4E-07 5.4E-08 Internal Fire CDF 4.2E-05 4.9E-05 7.1E-06 Seismic CDF 9.8E-07 1.9E-06 8.9E-07 Total Individual 4.8E-05 5.8E-05 1.1E-05 CDF OTMHM CDF 4.8E-05 5.9E-05 1.1E-05 1.3E-06 (ICCDP)

Internal Events LERF 3.1E-07 3.1E-07 5.2E-09 Internal Flooding 6.6E-09 8.0E-09 1.4E-09 LERF Internal Fire LERF 2.3E-06 2.6E-06 3.0E-07 Seismic LERF 2.5E-07 3.4E-07 9.1E-08 Total Individual 2.9E-06 3.3E-06 4.0E-07 LERF OTMHM (ICLERP)

LERF 3.0E-06 3.5E-06 4.5E-07 5.6E-08

Assessment of off-line modes Hatch does not have a quantitative process for evaluating low power and shutdown risk. It is noted that it is unnecessary to evaluate the low-power and shutdown contribution to the base CDF and

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LERF since the change being proposed involves performance of the repair while at-power. PSW provides motor cooling for RHR Service Wate r pumps during shutdown. Since RHRSW is used for shutdown cooling, and RHRSW is cooled by PSW, there is some risk involved with going into lower modes; however, that is not quantified or discussed any further in this assessment. The risk increase during shutdown is thus not zero and this is another reason for performing the repair on-line.

Compensatory Measures Discussion Risk Insights Risk insights from this configuration were ex amined by comparing the change in Fussell-Vesely (FV) values between the base and configuration specific importance rankings. The importance rankings were generated from the global all-hazards cutsets using CAFTA. A spreadsheet called BASE_PSW1A_Importance.xlsx contains the base model importance reports and the configuration specific importance reports. FV values that increased by a factor of three (200%) or more were examined to see what basic events contributed more to overall risk due to the failed pump. Because components and operator actions may be represented by multiple basic events, the overall risk increase is the sum of the individual event F-V terms. In the BASE_PSW1A_Importance.xlsx spreads heet, there are over 200 basi c events that increased by 200% or more. The events and components that become more important are associated with the other PSW pumps and Diesels 1B and 1C. These are discussed in more detail below.

Tier 2 Evaluation - Avoidance of risk significant configurations.

RG 1.177 requires an examination of other components that, in combination with the component already out of service, could result in a risk significant configuration. Maintenance Rule risk ranks components primarily on RAW, so that was c hosen for component measures. Time Critical Operator Action guidance from the PWROG ranks operator actions using Birnbaum, so that risk measure is also chosen. The BASE_PSW1A_Importance.xlsx spr eadsheet was used to identify basic events where the selected importance measure increased by a factor of three (delta risk greater than 200%) over the base case measure. These basic events and interpretation from the model are then used to help determine the Compensatory Measures.

For the PSW 1A pump, the items below had Fussell-Vesely risk increases greater than a factor of three.

Table 2-5 Risk-Significant Components Identified through Importance Measures Component Description

1R43S001B 1B Diesel Generator

1R43S001C 1C Diesel Generator

1P41C001B 1B PSW Pump

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1P41C001C 1C PSW Pump

1P41C001D 1D PSW Pump

Table 2-6 Risk-Significant Initiating Events Identified through Importance Measures

Initiating Component Description

Fire 0024A Cable Spreading Room

Fire 1408 4160V Bus Room 1F

Table 2-7 Risk-Significant Operator Actions Identified through Importance Measures

Operator Action ID Description

OPHERHRSWPMPCL Restore/Crosstie RHRSW pump motor cooling (34AB-P41-001-1).

(and -F, -L, -S versions)

OPHEPSWXTIE Crosstie Reactor Building Division 2 header to Division 1 header.

(and -F, -L, -S versions)

The above operator actions are associated with loss of service water to the RHRSW pumps and to the containment coolers, based on one of two pumps in division 1 already out of service.

Compensatory Measures Because the ICCDP is slightly above 1E-06, com pensatory measures are required. These measures are based on procedural protection, operation of redundant functions, and recommended actions based on the risk insights from the risk significant configurations above.

The following equipment is protected as required by SNC Procedure NMP-OS-010-002 for 1A PSW Pump out-of-service:

1C PSW Pump

1F 4160V Frame 4 (power supply to 1C PSW Pump)

1C PSW Pump Control Switch

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Per SNC procedure NMP-OS-010, Protected Equipment means:

Systems or components that have been identified as essential to ensuring that safety functions or unit generation is maintained for given plant conditions.

Components or equipment redundant to equipment that has failed, is taken out of service, degraded, or is otherwise unavailable when the redundant equipment is required for current plant conditions.

Physical barriers or signage are used to alert personnel to maintain a safe distance from the Protected Equipment in order to prevent unintended consequences from operation, maintenance, or nearby activity. It is noted that when prot ecting a system, SNC also protects systems supporting the primary system.

Additionally, the 1B and 1D PSW are being protected by Operations to mitigate the entry into another TS Action (TS 3.7.2 Action C). Plant Hatch Operations have initiated protecting the 1B, 1C, and 1D PSW pumps prior to the submittal of this LAR (NL-22-0316). This is reflected daily in the Shift Managers Briefing (Morning Report).

Based on the operator actions and equipment above, the following additional compensatory actions are recommended prior to ex ceeding the original 30-day LCO.

The operator actions are contained in procedure 34AB-P41-001-1 and an Operations Standing Order for shift briefing on the procedure associated with these actions will be in place prior to exceeding the original 30-day CT.

Section 4.1.7 Crosstie Cooling Sources This ensures that adequat e cooling will remain available for the RHRSW pump

Section 4.1.9 Operating Strategy for Event Mitigation Preparation for potential loss of an additional PSW pump

Protect the 1C diesel generator, 1B diesel generator and the Standby Service Water (SSW) pump 2P41C002 that provides normal cooling to the 1B DG and limit work to required TS surveillances only.

Work in the cable spreading room and the 1F switchgear room would be limited to operator rounds and surveillance activities.

As part of the uncertainty for the tornado hazard risk, site to do a proactive and preliminary walkdown and check for loose objects per site procedure 34AB-Y22-002-0.

Tier 3 Evaluation - A(4) Maintenance Rule configuration risk management impact.

Pump PSW 1A was input into the on-line configuration risk management (CRM) program. The Hatch CRM calculates both the instantaneous and integrated risk and CRM risk levels are based on integrated risk levels. The components already out of service prior to the 1A PSW Pump failure were left out of service for this evaluation to ensure the calculation is conservative. With the 1A PSW Pump out of service, the increase in risk is minimal as shown on Attachment 1. Because the CRM program uses the same hazard models that were used for this evaluation, and since the a(4) process evaluates planned work as well as current configurations, it will identify any potential

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high risk conditions and the a(4) process of assessing and managing that risk will adequately control the evolution. This includes any risk management actions that may occur during risk evaluations for rigging and lifting the motor and pump near other components that are remaining in service. The work schedules for the additional 15 day CT extension were also evaluated in the on-line CRM program to identify and re-schedule any potential risk-sign ificant planned work.

2.3 EXTERNAL EVENTS 2.3.1 Assessment of Relevant Hazard Groups Assessment of Relevant Hazard Groups The purpose of this portion of the assessment is to evaluate the spectrum of external event challenges to determine which external event hazards should be explicitly addressed as part of the TS 3.7.2 Condition A CT extension risk assessment.

A plant-specific evaluation of an extensive set of other external hazards (including high winds and external flooding) was performed in SNC calculation H-RIE-OEE-U00. The results have been previously submitted to the NRC for the Hatch 50.69 license amendment request (LAR) (ADAMS Accession Number ML18158A583) and subsequent RAI responses (ML19197A097). That evaluation was performed using the criteria in ASME PRA Standard RA-Sa-2009 and concluded that all other external hazards (i.e., all external hazards other than internal fires and seismic) can be screened from applicability at Hatch.

2.3.2 Other External Hazards Evaluation and Conclusions During the Hatch RICT LAR Audit, the screening criteria for the tornado missiles hazard was questioned. An RAI is expected from the Hatch RICT LAR Audit to discuss an updated screening criteria for beyond-design-basis tornado missiles. To address tornado missiles for this evaluation, the site already has a procedure that instructs walkdowns to be performed and ensure loose objects are secured if severe weather warnings are issues. This was further discussed in the compensatory measure recommendations. Also, a s ensitivity was done as part of this evaluation by increasing the LOSP initiating event in the PR A model. The results of the sensitivity showed minimum impact on delta risk. Therefore, there is no significant other external hazard risk contribution for this application.

2.4 RESULTS COMPARISON TO ACCEPTANCE GUIDELINES The results indicate a one-time extension up to 45 days would not exceed the ICCDP and ICLERP risk limits. Additional compensatory measures woul d potentially reduce risk further, such as protected equipment processes and other identified activities that impact potential to reduce LOOP probability. The additional compensatory measures are not accounted for in the quantification.

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2.5 UNCERTAINTY ASSESSMENT The purpose of this section is to disposition th e impact of Probabilistic Risk Assessment (PRA) modeling epistemic uncertainty for Condition 3.7.2 Required Action A.1 CT extension assessment. Assumptions and Uncertainties for the Base Hazard Calculations The purpose of this section is to disposition the impact of Probabilistic Risk Assessment (PRA) modeling epistemic uncertainty for Condition 3.7.2 Required Action A.1 CT extension assessment. The baseline internal events PRA, internal flooding PRA, fire PRA (FPRA) and seismic PRA models document assumptions and sources of uncertainty and these were reviewed during the model peer reviews. The approach taken is, therefore, to review these documents to identify any items which may be directly relevant to Condition 3.7.2 Required Action A.1 CT extension assessment, discuss the results, and to provide dispositions. No key assumptions or sources of uncertainty were identified that uniquely impact this application.

Several additional areas of uncertainty were also investigated.

1) The site has discovered several circuit breakers that are not well coordinated with downstream loads. The listing of breaker and loads was reviewed, and none of them impact the PSW pumps or the components that become more important with a pump out of service. The Hatch Fire PRA was built with the assumption that all credited power supplies in the Fire PRA were coordinated with their upstream breakers. After further review, it was identified that there is a subset of breakers that are in fact not coordinated with the upstream breakers. PRA calculation, PRA-BC-H-21-006, documents the initial review of the impact of these uncoordinated breakers.

This calculation evaluated these breakers in a conservative manner in that each breaker was failed in every scenario. The calculation also added additional detailed circuit analysis to the Hatch Fire PRA model. The conclusion on the calculation determined only two equipment that needed further analysis. The coordination issue only exists when they are on the alternate feed.

The current alignment during this evaluation have both pieces of equipment aligned via their normal feed. Therefore, the uncoordinated breakers will have a minimum impact on the 1A PSW Pump evaluation.

2) For tornado missile hazard, Loss of Offsite Power is one of the initiating events that becomes more important with a PSW pump out of service. This is addressed by a sensitivity run that increases the weather portion of the LOSP initiator to its 95% value. This value was calculated previously for the Unit 1 Diesel Ge nerator liner replacement LCO extension in SNC calculation PRA-BC-H-20-001.

3) The treatment of open phase protection (OPP) of the startup transformers is a potential source of uncertainty. Hatch did not add OPP equipment or initiating events to the PRA models, based on the event screening done in accordance with NEI guidance. Because of the arrangement of offsite power sources, three startup transformers per unit supplying three 4160v switchgear, an OPC would not cause a plant trip. Additionally, at Hatch the OPIS panels are alarm only and operator action is required to isolate impacted equipment. The impact of an undetected OPC would be equivalent to the loss of a 4KV bus, which is modeled. The probabilities associated with the OPC events, detection equipment failures, and operator action failures, results in the OPC contribution to risk being several orders of magnitude smaller than

Attachment 4, Page 21 NL-22-0316, Attachment 4 Evaluation of Risk Impact and Compensatory Measures

the random failures of the startup transformer s or 4160v switchgear. The impact of not modeling the OPC condition and mitigating equipment and actions is very low. This does not represent a significant source of uncertainty for evaluating PSW pumps. This evaluation is contained in the Hatch OPP evaluation calculation PRA-BC-H-19-004.

Sensitivities

Tornado impacts.

Calculation PRA-BC-H-20-001 performed for the Unit 1 Diesel Generator LAR extensions calculated an increase in the %IE-LOSP initiating event frequency using the 95% value for the weather portion. This increases the initiator frequency from 2.115E-02/yr. to 3.378E-02/yr. The existing PSW1A cutset for Internal Events and OTMHM was used by increasing the frequency.

This changed the OTMHM CDF risk from 5.9E-05/yr. to 6.1E-05/yr. This is reflected below. The OTMHM LERF risk had a very small increase from 3.47E-06/yr. to 3.48E-06/yr.

Table 2-8 Increased LOSP Initiating Event Sensitivity Integrated Value for Hazard Base risk (/yr.) New risk (/yr.) Delta Risk (/yr.) 15-day extension (45-day total CT)

Internal Events CDF 4.1E-06 8.6E-06 4.5E-06 Internal Flooding CDF 2.9E-07 3.4E-07 5.4E-08 Internal Fire CDF 4.2E-05 4.9E-05 7.1E-06 Seismic CDF 9.8E-07 1.9E-06 8.9E-07 Total Individual CDF 4.8E-05 6.0E-05 1.3E-05 OTMHM 4.8E-05 6.1E-05 1.3E-05 1.6E-06 (ICCDP)

Internal Events LERF 3.1E-07 3.2E-07 1.4E-08 Internal Flooding LERF 6.6E-09 8.0E-09 1.4E-09 Internal Fire LERF 2.3E-06 2.6E-06 3.0E-07 Seismic LERF 2.5E-07 3.4E-07 9.1E-08 Total Individual LERF 2.9E-06 3.3E-06 4.1E-07 OTMHM 3.0E-06 3.5E-06 4.6E-07 5.7E-08 (ICLERP)

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These risk increases are small and still within regulatory guidance.

3.0 TECHNIC AL ADEQUACY O F PRA MODEL This section provides information on the technical adequacy of the Hatch Nuclear Plant Probabilistic Risk Assessment (PRA) models. The Hatch PRA maintenance and update processes and technical capability evaluations prov ide a robust basis for concluding that the PRA is suitable for use in risk-informed licensing actions, specifically in support of the requested extended CT for TS Condition 3.7.2 Required Action A.1. The One-Top Multihazard Model (OTMHM) is comprised of various hazards PRA models that can be quantified simultaneously or individually. Each hazard (internal events, internal flooding, fire, and seismic) has been peer reviewed. The most up-to-date assessments of PRA technical adequacy (including peer review status, F&O closure status, scope, fidelity, capability, and maintenance/update practices) was provided to the NRC previously for the Hatch 50.69 LAR (ADAMS Accession Number ML18158A583) and subsequent RAI responses (ML19197A097); and also the NFPA-805 LAR (ML18096A955) and subsequent RAI responses (ML19280C812). Additionally, those submittals contain the most up-to-date description of the other external hazards assessment.

4.0

SUMMARY

AND CONCLUSIONS This analysis evaluates the accept ability, from a risk perspective, of a change to the Hatch Unit 1 TS Condition 3.7.2 Required Acti on A.1 for a one-time increase of the CT from 30 days to 45 days when the 1A PSW Pump is inoperable.

The analysis examines a range of risk contributors including internal events, internal flooding, fire, seismic, shutdown risk and other external hazards. The configuration was quantified using the Phoenix OTMHM model and compared to the base risk to obtain delta CDF and LERF values.

4.1 PRA QUALITY The PRA quality has been assessed and determined to be adequate for this risk application, and the PRA technical adequacy has also been addressed in recent NRC submittals.

To summarize,

Scope - Hatch PRA modeling is highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause events. The PRA has the necessary scope to appropriately assess the pertinent risk contributors.

Fidelity - The Hatch PRA models are the most recent evaluation of the risk profile.

The PRA reflects the as-built, as-operated plant, with the exception of previously noted items.

Standards - The PRA has been reviewed against the ASME /ANS PRA Standard and the PRA elements are shown to have the necessary attributes to assess risk for this application.

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Peer Review - The PRA has received a peer review. Based on addressing the peer review results and subsequent gap analyses to the current standards, the PRA is found to have the necessary attributes to assess risk for this application.

Appropriate Quality - The PRA quality is found to be appropriate to assess risk for this application.

4.2 QUANTITATIVE RESULTS VS. ACCEPTANCE GUIDELINES This analysis demonstrates with reasonable assurance that the proposed TS change is within the current risk acceptance guidelines in RG 1.177 for one-time changes. A d d i t i o n a l s e n s i t i v i t y analysis show that the RG 1.177 thresholds are not challenged.

4.3 CONCLUSION

S This analysis demonstrates the acceptability, from a risk perspective, of a change to the Hatch TS Condition 3.7.2 Required Action A.1 to increase the CT from 30 days to 4 5 days when the 1A PSW Pump is unavailable.

A PRA technical adequacy evaluation was also performed consistent with the requirements of ASME/ANS PRA Standard and RG 1.200, Revision 2. Additionally, a review of model uncertainty and outstanding changes was performed with this application. None of the identified sources of uncertainty were significant enough to change the conclusions from the risk assessment results presented here.

5.0 REFERENCES

[1] Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities, Revision 2, December 2020.

[2] Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,"

Revision 3, January 2018.

[3] Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," Revision 2, January 2021.

[4] ASME/ANS RA-Sa-2009, February 2009. Addenda to RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applicati ons,

[5] H-RIE-PHOENIX-U01, One Top Model for PHOENIX Configuration Risk Management Program.

[6] H-RIE-OEE-U00 - Hatch Other External Events Screening

[7] RBA-22-007-H - PSW 1A Emergent Technical Specification Change

Attachment 4, Page 24

v t e Letter Sequence Request
EPID:L-2022-LLA-0062, (Withdrawn, Closed)
Initiation Request Acceptance... Administration Questions Answers Results Withdrawal Other: ML22123A053
MONTHYEAR2022-11-30 [Table View]

NL-22-0316, Emergency License Amendment Request for Technical Specification 3.7.2 Regarding One-Time Extension of Completion Time for Plant Service Water (Psw) Pump Inoperable

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Enclosure:

1.0 INTRODUCTION

1.2 BACKGROUND

SUMMARY

4.3 CONCLUSION

5.0 REFERENCES

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NL-22-0316, Emergency License Amendment Request for Technical Specification 3.7.2 Regarding One-Time Extension of Completion Time for Plant Service Water (Psw) Pump Inoperable

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Enclosure:

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1.2 BACKGROUND

SUMMARY

4.3 CONCLUSION

5.0 REFERENCES

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