View source for L-2022-176, License Amendment Request to Revise Cooling Tower Service Water Loop or Cell Requirements

{{Adams
| number = ML22343A259
| issue date = 12/09/2022
| title = License Amendment Request to Revise Cooling Tower Service Water Loop or Cell Requirements
| author name = Strand D
| author affiliation = NextEra Energy Seabrook, LLC
| addressee name = 
| addressee affiliation = NRC/NRR, NRC/Document Control Desk
| docket = 05000443
| license number = NPF-086
| contact person = 
| case reference number = L-2022-176
| document type = Letter type:L, License-Application for Facility Operating License (Amend/Renewal) DKT 50
| page count = 1
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:December 9, 2022 L-2022-176 10 CFR 50.90

U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington DC 20555-0001

RE: Seabrook Station Docket No. 50 -443 Renewed Facility Operating License No. NPF-86

License Amendment Request to Revise Cooling Tower Service Water Loop or Cell Requirements

Pursuant to 10 CFR 50.90, NextEra Energy Seabrook, LLC (NextEra) hereby requests an amendment to Renewed Facility Operating License NPF-86 for Seabrook Nuclear Plant Unit 1 (Seabrook). The proposed license amendment modifies Seabrook Technical Specifications (TS) 3/4.7.4, Service Water System / Ultimate Heat Sink, by increasing the allowable outage time (AOT) for an inoperable cooling tower service water loop or cell. Additionally, the proposed license amendment would make an editorial correction to TS Section 1.9.

The enclosure to this letter provides NextEra's evaluation of the proposed changes. Attachment 1 to the enclosure provides the Seabrook TS pages marked up to show the proposed changes. Attachment 2 provides the Seabrook TS Bases pages marked up to show the proposed changes. Attachment 3 provides a risk-informed analysis supporting the proposed change based on the current Seabrook probabilistic risk assessment (PRA). The TS Bases changes are provided for information only and will be incorporated in accordance with TS Bases Control Program upon implementation of the approved amendments.

NextEra has determined that the proposed license amendment does not involve a significant hazards consideration pursuant to 1 O CFR 50.92(c), and that there are no significant environmental impacts associated with the change. The Seabrook Onsite Review Group (ORG) has reviewed the enclosed amendment request. In accordance with 10 CFR 50.91(b)(1), a copy of this license amendment request is being forwarded to the designee for the State of_ New Hampshire.

NextEra requests that the proposed license amendment is processed as a normal license amendment request, with approval within one year and implementation within the following 90 days.

This letter contains no new regulatory commitments.

Should you have any questions regarding this submission, please contact Mr. Kenneth Mack, Licensing Manager at (561) 904 - 3635.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on the q"tl\\.day of December 2022.

Sincerely, 11{ ~ s1=-==--------
ianne Strand General Manager Regulatory Affairs

Enclosure Attachments

NextEra Energy Seabrook, LLC P.O. Box 300, Lafayette Road, Seabrook, NH 03874 Seabrook Station L-2022-176 Docket Nos. 50-443 Page 2 of 2

cc: USNRC Region I Administrator USNRC Project Manager USNRC Senior Resident Inspector

Director Homeland Security and Emergency Management New Hampshire Department of Safety Division of Homeland Security and Emergency Management Bureau of Emergency Management 33 Hazen Drive Concord, NH 03305

Katharine Cederberg, Lead Nuclear Planner The Commonwealth of Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702-5399 Seabrook Station L-2022-176 Docket Nos. 50-443 Enclosure Page 1 of 37

Evaluation of the Proposed Changes

Seabrook Station

License Amendment Request to Revise Cooling Tower Cell Requirements

1.0 

==SUMMARY==
DESCRIPTION............................................................................................................. 2

2.0 DETAILED DESCRIPTION.............................................................................................................. 2

3.0 TECHNICAL EVALUATION............................................................................................................ 4

==4.0 REGULATORY EVALUATION==
...................................................................................................... 10

5.0 ENVIRONMENTAL CONSIDERATION......................................................................................... 13

==6.0 REFERENCES==
............................................................................................................................... 13

ATTACHMENTS
: 1. Technical Specifications pages (markup) 2. Technical Specifications Bases pages (markup)
: 3. Risk-based Supporting Analysis Seabrook Station L-2022-176 Docket Nos. 50-443 Enclosure Page 2 of 37

1.0 

==SUMMARY==
DESCRIPTION

Pursuant to 10 CFR 50.90, NextEra Energy Seabrook, LLC (NextEra) requests an amendment to Renewed Facility Operating License NPF-86 for Seabrook Nuclear Plant Unit 1 (Seabrook). The proposed license amendment modifies Seabrook Technical Specifications (TS) 3/4.7.4, Service Water System / Ultimate Heat Sink, by increasing the allowable outage time (AOT) for one inoperable cooling tower service water loop or one cooling tower cell. Additionally, the proposed license amendment would make an editorial correction to TS Section 1.9.

2.0 DETAILED DESCRIPTION

2.1 System Design and Operation

The ultimate heat sink complex consists of the Atlantic Ocean and atmosphere. The Atlantic Ocean serves as the normal ultimate heat sink for Seabrook Station. In the unlikely event that the normal supply of cooling water from the Atlantic Ocean is unavailable, the atmosphere serves as the ultimate heat sink through the use of a mechanical draft evaporative cooling tower.

The Atlantic Ocean portion of the ultimate heat sink is designed to perform all safety functions during and following the most severe natural phenomena anticipated, e.g., the safe shutdown earthquake (SSE), tornado, hurricane, flood, or low water level resulting from storm surges with the exception of the tunnels and transition structure, which are not designed for the SSE. In the unlikely event that an earthquake of sufficient intensity occurs, which blocks over 95 percent of the flow area of the intake tunnel, the cooling tower would be used as the ultimate heat sink.

In the unlikely event that the main circulating water tunnel is unavailable, a mechanical draft evaporative cooling tower serves as the ultimate heat sink. The cooling tower is designed to supply cooling water to the primary component cooling water and diesel heat exchangers while sustaining a loss of offsite power and any single active failure. The tower, tower pumps and all its associated components are designed for the SSE loads, which assures that cooling water will be available from the ultimate heat sink complex during and following all natural phenomena.

The mechanical draft cooling tower provides an alternate source of cooling water th.at is completely independent of the circulating water tunnels and the Atlantic Ocean. In the unlikely event that level is lost in the Service Water Pumphouse, the heat loads are transferred from the Atlantic Ocean to the cooling tower.

The cooling tower can be used during normal operation subject to the level and temperature limitations listed in TS. In the unlikely event that the cooling tower is unavailable, the Atlantic Ocean will supply cooling to components.

Transfer of heat loads to the cooling tower can be performed manually on a system or component level from the main control board. Automatic transfer will occur on a train basis upon actuation of the associated tower actuation logic. The tower actuation logic, associated with each train, senses service water pump discharge pressure. A low pressure condition, indicative of a low-low service water pumphouse level, will initiate tower actuation for the associated service water train.

The cooling tower complex is that portion of the ultimate heat sink that includes a three (two of which are functional) cell tower, a basin with five (three of which are functional) interconnected compartments, two (one of which is functional) pump rooms and the Seabrook Station L-2022-176 Docket Nos. 50-443 Enclosure Page 3 of 37

associated piping, valves, and equ ipment. The functional portion of the cooling tower consists of one independent cell with one fan and a center cell with two fans. The funGtional portion of the basin consists of a pump well and one catch basin for each of the two functional tower spray cells. The unit has an "A" and a "B" cooling tower complex flow train. The pumps with associated piping, valves, and equipment in the trains circulate cooling water from the pump well basin through the primary component cooling heat exchangers and the secondary component cooling heat exchangers during normal operations or the diesel generator heat exchangers during loss of offsite power conditions or both during test. The flow is returned to the basin through either the respective tower sprays or through the spray bypass header, which distributes the return flow to each of the two tower cell catch basins. A heat exchanger bypass line is provided from each pump discharge to the return line permitting cooling tower spray or spray bypass header recirculation independent of normal ocean cooling operations.

Each train-associated tower pump, fan, and associated electrical equipment serving a single primary component cooling water heat exchanger has a common emergency electrical power supply and is separated from the other train's power supply. A loss of power to the electrical equipment supplying one flow train would affect only that flow train and would still allow sufficient capacity for cooling the unit under a LOCA condition.

2.2 Current Requirements/ Proposed Changes to TS 3/4.7.4

In Modes 1 through 4, TS 3/4.7.4, Service Water System/ Ultimate Heat Sink, requires an Operable mechanical draft cooling tower and two cooling tower service water loops w ith one Operable cooling tower service water pump in each loop.

TS 3.7.4, ACTION b stipulates:

 "With one cooling tower service water loop or one cooling tower cell inoperable, return the affected loop or cell to OPERABLE status within 7 days, or be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours."

The proposed change extends from 7 days to 21 days the AOT to return the inoperable cooling tower service water loop or cell to Operable status.

2.3 Reason for the Proposed Changes to TS 3/4.7.4

The proposed license amendment is needed to provide sufficient time for cooling tower service water loop or cell troubleshooting and maintenance without initiating a plant shutdown or requesting Exercise of Enforcement Discretion and Emergency License Amendment to avert an orderly shutdown within 7 days.

2.4, Editorial Correction to TS Section 1.9

TS Section 1.9 states:

CORE ALTERATION shall be the movement or any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE AL TERATIONs shall not preclude completion of movement of a component to a safe position.

TS Section 1.9 would be revised to state:
Seabrook Station L-2022-176 Docket Nos. 50-443 Enclosure Page4 of 37

CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONs shall not preclude completion of movement of a component to a safe position.

2.5 Reason for Editorial Correction to TS Section 1.9

Reference 6.6 issued Amendment No. 81 for changes to the definitions in TS Sections 1.9, Core Alteration, 1.14, Engineered Safety Features Response Time, and 1.29 Reactor Trip Response Time. TS page 1-2 issued with Amendment No. 81 revised TS Section 1.9 to read, "CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONs shall not preclude completion of movement of a component to a safe position."

Reference 6.7 requested changes to Turkey Point Units 3 and 4, Seabrook Station, and Point Beach Units 1 and 2 TS definitions to adopt TSTF-563, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program. Reference 6.7 requested changes to Seabrook TS page 1-1 to change Sections 1.3, Analog, Channel Operational Test and 1.5, Channel Calibration; TS page 1-2 to change Section 1. 11, Digital Channel Operational Test; and TS page 1-7 to change Section 1.39, Trip Actuating Device Operational Test. The marked-up TS page 1-2 that proposed a change to TS Section 1.11 included TS Section 1.9 that wasn't marked-up, but read, "CORE ALTERATION shall be the movement or any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE AL TERATIONs shall not preclude completion of movement of a component to a safe position." Reference 6. 7 inadvertently proposed a change to TS Section 1.9.

Reference 6.8 issued Amendments to Point Beach Units 1 and 2, Seabrook Station, and Turkey Point Units 3 and 4 that revised the TS to adopt TSTF-563. Included in Reference 6.8 was Seabrook Amendment 164 that issued TS page 1-2 that revised TS Section 1.11, but also included the inadvertent change to TS Section 1.9 that read, "CORE ALTERATION shall be the movement or any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONs shall not preclude completion of movement of a component to a safe position."

3.0 TECHNICAL EVALUATION

The proposed change to TS Section 1.9 is editorial in nature and does not require a technical evaluation.

3.1 Increase the Allowable Outage Time (AOT) for One Inoperable Cooling Tower Service Water Loop or One Inoperable Cooling Tower Cell

The proposed license amendment modifies Seabrook TS 3.7.4, ACTION b, by changing the AOT from 7-days to 21-days.

In evaluating the proposed AOT extension, NextEra applied Regulatory Guide (RG) 1.177, "Plant-Specific, Risk-Informed Decision-making: Technical Specifications,"
(Reference 6.4). RG 1.177 describes acceptable methods for assessing the nature and impact of proposed TS changes by considering engineering issues and applying risk insights. The approach provides for probabilistic risk assessment (PRA) state-of-the-art methods in a manner that complements deterministic considerations and traditional Seabrook Station L-2022-176 Docket Nos. 50-443 Enclosure Page 5 of 37

defense-in-depth philosophy, consistent with Nuclear Regulatory Commission's (NRC's) policy "Use of Probabilistic Risk Assessment Methods in Nuclear Activities: Final Policy Statement," (Reference 6.5). RG 1.177 establishes a three-tiered approach to licensee evaluation of the risk associated with AOT changes. Tier 1 evaluates the risk impact expressed as changes to the core damage frequency (b.CDF), incremental conditional core damage probability (ICCDP), large early release frequency {b.LERF), and incremental conditional large early release probability (ICLERP). Tier 2 evaluates the dominant-risk plant configurations to assure appropriate restrictions will be in place. Tier 3 evaluates the licensee's overall configuration risk management program (CRMP) to assure potentially risk-significant configurations are adequately managed. Tiers 1, 2 and 3 are evaluated by addressing each of the Engineering Evaluation elements of RG 1. 177 described below.

3.1.1 Regulatory Compliance

No exceptions or exemptions from applicable regulations or accepted industry codes and standards relevant to safe operation are proposed. As a part of the Seabrook Service Water System/ Ultimate Heat Sink, the cooling tower cells satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii) for TS inclusion as an LCO. The proposed AOT extension would not contravene compliance with the applicable ACTION(s) and surveillance requirements (SRs) or challenge the Service Water System/ Ultimate Heat Sink capability to function as described in Criterion 3. In the event an inoperable cooling tower service water loop or cell cannot be restored within the proposed AOT extension, the Seabrook TS require that the LCO be considered not met and the appropriate ACTION must be entered (i.e.,
be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours). In this scenario, the plant would proceed with an orderly shutdown in the same manner as the current AOT requirement. Thereby, cooling tower service water loop or cell capability to function consistent with applicable requirements and safety analysis assumptions is unaffected by the proposed change.

3.1.2 Traditional Engineering Considerations

3.1.2.1 Defense in Depth

During the proposed AOT extension, defense - in-depth measures will be applied to account for unknown and unforeseen failure mechanisms or other phenomena to assure the safety function of the Service Water System/ Ultimate Heat Sink is maintained. This includes the NextEra online risk management process to assess the impact of the inoperability and undertake appropriate maintenance repair or replacement activities to minimize risk. As discussed in Sections 3.4 and 3.6 of this amendment request, risk-significant plant configurations will not be entered and risk-reduction measures will be implemented to maintain defense in depth. By implementing the multiple, independent and redundant layers of defense as summarized below, the integrity of barriers to core damage will be maintained.

o The proposed AOT extension does not affect the balance among the core damage prevention, containment fa ilure prevention and consequence mitigation principles. During the proposed AOT extension, supported equipment such as the emergency diesel generator and primary component cooling water system are not rendered inoperable such that the assumptions and inputs associated with plant safety analyses are unaffected. Thereby, the balance of prevention and mitigation strategies remains preserved.
Seabrook Station L-2022-176 Docket Nos. 50-443 Enclosure Page 6 of 37

o The proposed AOT extension does not create an over-reliance on existing programmatic activities as compensatory measures. Station response to an inoperable cool ing tower service water loop or cell beg ins with entering the appropriate ACTION and evaluating the risk-significance of the repair consistent with 1 O CFR 50.65(a)(4). Extend ing the AOT neither modifies the conditions warranting ACTION entry nor the risk-based considerations and station activities which assure safe operation.

o The proposed AOT extension maintains the redundancy, independence and diversity of systems commensurate with the expected frequency and consequences of system challenges. As demonstrated in Section 3.3 of this amendment request, the risk associated with the extended AOT duration is sufficiently low.
* Station response to concurrent equipment inoperability is unchanged by the proposed change, including cessation of the ma intenance or plant shutdown if wa rranted. The extended AOT neither increases the likelihood nor the consequences of simultaneous equipment malfunctions since one of the two train-related cooling tower service water loops or cells will remain Operable and fully functional during the extended AOT. Should simultaneous equ ipment outages occur, the online risk management process will evaluate and implement appropriate risk-reduction measures.
* Compensatory actions to be taken when entering the extended AOT will be promptly identified and implemented as appropriate for managing the risk associated with the repair consistent with 10 CFR 50.65(a)(4).
* The station's online risk management process will continue to evaluate planned maintenance repair for risk-significant configurations, concurrent equipment outages, abnormal plant conditions, and external events such as challenges to grid stability and adverse weather conditions. The proposed AOT extension does not alter the manner in which these considerations are factored into the online risk assessment process.
* During the proposed AOT extension, one of the two train-related cooling tower service water loops or cells will remain Operable and fully functional. As such, no disruption to the safety function of the Service Water System I Ultimate Heat Sink will occur during the proposed AOT extension. All safety analysis assumptions and inputs remain valid.

o The proposed AOT extension cannot reduce the defenses against or increase the likelihood of a common-cause failure (CCF) or introduce new CCF mechanisms. In the event of an inoperable cooling tower service water loop or cell, the redundant coo ling tower service water loop or cell is sufficiently instrumented and monitored such that any CCF would be qu ickly identified and appropriate action promptly taken. When a non-conforming or degraded condition is identified, the process of evaluating operability, conducted by a licensed senior reactor operator, assesses the potential for common causes and effects on the other trains and components. If a common cause issue is present, it will be accounted for in the operability determination prior to determining the appropriate ACTION to be entered.
No changes are proposed to plant equipment or the manner in which equipment is evaluated for operability, including consideration for CCFs.
Seabrook Station L-2022-176 Docket Nos. 50-443 Enclosure Page 7 of 37

o The proposed AOT extension does not alter any guarded equipment practices which protect vital equipment or equipment tagg ing practices designed to enhance personnel safety. No new deviations or exceptions are proposed to the methods of establishing and maintaining physical equipment barriers during the repair such that barrier independence would be degraded.

o During the proposed AOT extension, human performance practices such as pre-job briefs, job site reviews, place-keeping, etc., which reduce the likelihood of human errors will continue to be implemented in accordance with plant administrative and implementing procedures. The proposed AOT extension only extends the time the inoperable cool ing tower service water loop or cell can be out of service without initiating a plant shutdown.
Extending the AOT cannot lessen the defenses against human errors implemented through plant procedures.

o The proposed AOT extension concerns a maintenance act ivity and thereby cannot alter the intent of any plant or equipment design criteria. The proposed change provides for orderly maintenance which restores an inoperable cooling tower service water loop or cell to its plant design as currently licensed using authorized maintenance practices. Any changes to the cooling tower service water loop or cell, including performance criteria, setpoints, etc., is subject to screening for prior NRC approval in accordance with 10 CFR 50.59.
: 3. 1.2.2 Safety Margin

During the proposed AOT extension, the unaffected cooling tower service water loop or cell would remain fully Operable and capable of performing its specified function. As such, the proposed AOT extension does not affect equipment functions, response times or acceptance criteria, do not introduce new or altered methods of assessing plant performance, and do not alter the manner in which the station would respond to a concurrent equipment malfunction. All safety analysis assumptions and inputs are unaffected and the marg in to plant safety limits and limiting safety settings are unchanged. As such, there is no reduction in the margin to safety as a result of the proposed change.

3.3 Evaluation of Risk Impact

Attachment 3 of this amendment request provides the evaluation of risk impact for the proposed AOT extension, including a discussion of PRA scope, technical adequacy, modeling and insights. The evaluation determined that the ICCDP and the ICLERP for the proposed AOT extension are below the RG 1. 177 threshold of 1.0E-6 per year ICCDP and 1.0E-07 ICLERP, respectively. The evaluation demonstrated that the increase in plant risk associated with extending the AOT from 7-days to 21-days for one cooling tower service water loop or cell inoperable is acceptably small.

3.4 Acceptance Guidelines for Technical Specification Changes

The proposed AOT extension for one cooling tower cell inoperable would be a permanent change to the Seabrook TS. The RG 1. 177 guidance for permanent AOT changes is consistent with the fundamental principle that changes to TS result in small increases in overall risk to the health and safety of the public. To assure this principal is satisfied, RG
: 1. 177 establishes a three-tiered approach to licensee evaluation of permanent AOT changes from a risk-based perspective, as addressed below :
Seabrook Station L-2022-176 Docket Nos. 50-443 Enclosure Page 8 of 37
: 1. The licensee has demonstrated that the TS completion time (CT) change has only a small quantitative impact on plant risk. An ICCDP of Jess than 1.0x10- 6 and an ICLERP of Jess than 1.0x10- 7 are considered small for a single TS condition entry (Tier 1).

The PRA analysis summarized in Section 3.3 and Attachment 3 demonstrates that the increase in the ICCDP and ICLERP associated with the proposed AOT extension has only a small quantitative impact on plant risk, thereby satisfying the Tier 1 criteria established in RG 1.177 for a permanent AOT change.
: 2. The licensee has demonstrated that there are appropriate restrictions on dominant risk-significant configurations associated with the change (Tier 2).

The PRA analysis summarized in Section 3.3 and Attachment 3 did not identify equipment outages or plant configurations having extremely high-risk contributions as a result of the proposed AOT extension. As such, no plant configurations or equipment outages require enhancement to the Seabrook TS or to plant programs and procedures. However, NextEra chooses to implement two qualitative, prudent compensatory measures that improve the Seabrook's defense in depth during the proposed AOT extension and further increases the available margin to acceptance guidelines, as indicated below.
The compensatory measures are not credited in the PRA analysis summarized in Section 3.3 and Attachment 3, but the actions will be implemented in recognition that during the period of one cooling tower cell being inoperable, the reliability of the Service Water System/ Ultimate Heat Sink is reduced while reliant on the one Operable cooling tower service water loop or cell and its associated Emergency Diesel Generator (EOG) in the event of a loss of offsite power. These voluntary compensatory measures are:
: 1) Entry into a proposed AOT extension will not be planned concurrent with EOG maintenance.
: 2) Entry into a proposed AOT extension will not be planned concurrent with maintenance on other RPS, EFSAS or containment isolation actuation instrumentation channels that could result in an affected channel being placed in a tripped condition.

The absence of dominant risk-significant configurations associated with the proposed AOT extension and the voluntary compensatory measures which assure appropriate restrictions against dominant risk-significant configurations satisfy the Tier 2 criteria established in RG 1.177 for a permanent AOT change.
: 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).

During the proposed AOT extension, any repair activities on the inoperable cooling tower service water loop or cell would first be evaluated for aggregate risk impacts to the station using NextEra's Risk Management Program and work activity risk management (WARM) procedures. These procedures are employed to evaluate, plan and manage equipment maintenance activities Seabrook Station L-2022-176 Docket Nos. 50-443 Enclosure Page 9 of 37

and include an assessment of risk associated with unavailable equipment as required by 10 CFR 50.65(a)(4). The evaluations provide a forward-looking assessment of potentially risk-significant activities warranting reductions to acceptable levels (i.e., low-risk whenever feasible). Significant risk activities are those having the potential to affect personnel, nuclear or radiological safety, environmental regulations, or power generation. Enhanced preparation, execution and oversight are required for each risk category on a graded scale.

Additionally, an online aggregate risk assessment is performed each shift in accordance with NextEra's Online Risk Management (OLRM) procedures.
The assessment employs probabilistic safety analysis calculations for online maintenance, safety train analyses to assure adequate separation and redundancy, and consideration of environmental factors such as severe weather and other challenges to grid stability. The online aggregate risk assessment applies the same fault trees and database as Seabrook's PRA model, and so is fully capable of evaluating changes in CDF and LERF for internal events. Should conditions change that challenge the maintenance repair, such as other equipment malfunctions, the on line risk determination would be re-performed to maintain an accurate risk profile for the most limiting plant condition during the current shift. Th~ profile update would include consideration for entry into applicable ACTIONs and for specific measures necessary to reduce the overall aggregate risk up to and including aborting the maintenance in progress. Medium aggregate risk activities and above would prompt work schedule changes which reduce aggregate risk.
High aggregate risk (HRA) additionally requires Operations Director approval along with resource commitments, enhanced protective measures and risk reduction contingencies and controls.

The work management and OLRM processes established in plant procedures ensure plant conditions are taken into account contemporaneously, equipment credited for supporting safe operation are protected, detailed pre-job briefings and job-site reviews are conducted, and that all parties are apprised of the risk impact(s) and their roles and responsibilities prior to mobilizing to perform the maintenance activity. The culmination of these activities is to conduct reviews and evaluations of work schedules before beginning work, determine the safety implications for performance, and assess, monitor and maintain acceptable levels of on-line risk and thereby satisfy the Tier 3 criteria established in RG 1.177 for a permanent AOT change.

3.5 Comparison of Risk of Available Alternatives

Alternatives to the proposed AOT extension have been considered during cooling tower service water loop or eel~ failures such as the preparation of NOEDs and exiting the applicable MODE. Preparation of a NOED is not an appropriate consideration for the risk based analysis summarized in Section 3.3 and Attachment 3 since the outcome from a risk-based perspective would be identical to the proposed change. A unit shutdown was not considered in the risk-based analysis due to the inherent increase in risk associated with shutdown evolutions. NextEra concludes that commencing a unit shutdown within 7-days of one cooling tower service water loop or cell being inoperable has greater safety implications than continuing power operation for the extended duration of time needed for repair of one cooling tower service water loop or cell given the redundancy in the Operable cooling tower service water loop or cell. In all cases, efforts to minimize the period of cooling tower service water loop or cell inoperability while maintaining Seabrook Station L-2022 - 176 Docket Nos. 50-443 Enclosure Page 10 of 37

acceptable risk at-power would be the primary focus in contrast to the additional challenges associated with the commencement of a unit shutdown within 7-days, and thereby would better suit overall safety.

3.6 Conclusion

The proposed AOT extension was evaluated against the deterministic and risk-based considerations presented in RG 1.177, including predicted changes to the CDF and LERF, consistent with the fundamental principle that any increase in risk to the health and safety of the public resulting from the change shall be negligible. The evaluation found that the proposed AOT extension satisfies the deterministic considerations for regulatory compliance, defense-in-depth and safety margin. The evaluation further determined that the quantitative impact on plant risk is small (Tier 1 ), that there are no dominant risk-significant configurations associated with the change (Tier 2) and that procedures to utilize, maintain, and control NextEra ' s risk-informed plant configuration control program are in place at Seabrook (Tier 3). Thereby, the proposed changes extending the AOT from 7-days to 21-days for one cooling tower service water loop or cell inoperable satisfies the RG 1.177 criteria for acceptability.

==4.0 REGULATORY EVALUATION==
4. 1 Applicable Regulatory Requirements/Criteria
* 10 CFR 50.36(c)(2)(i) states that when a limiting condition 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.
* 10 CFR 50.65, states, in part, that preventive maintenance activities must be sufficient to provide reasonable assurance that SSCs are capable of fulfilling their intended functions.
* GDC 19 of 10 CFR 50, Appendix A, states that the control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, fo r the duration of the accident. Equipment at appropriate locations outside the control room shall be prov ided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.
* Regulatory Guide (RG) 1.177 describes methods acceptable to the NRC staff for assessing the nature and impact of proposed TS changes by considering engineering issues and applying risk insights.
* Regulatory Guide 1.200 describes one acceptable approach for determining whether a base probabilistic risk assessment (PRA), in total or the portions 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.
Seabrook Station L-2022-176 Docket Nos. 50-443 Enclosure Page 11 of 37

The proposed license amendment complies with the requirements of 10 CFR 50.36(c),
10 CFR 50.65(a)(4), GDC 19 of 10 CFR 50, Appendix A, and RGs 1.177, Revision 2, and 1.200, Revision 2. All regulatory requirements and applicable guidance will continue to be satisfied as a result of the proposed license amendment.

4.2 Precedents

4.2.1 In Reference 6. 1, NextEra Energy Seabrook requested approval for enforcement discretion to allow delaying taking actions to provide for the necessary time to restore the service water cooling tower 51 B fan to an Operable status. On September 23, 2001, the NRC approved the requested NOED for issuance.

4.2.2 In Reference 6.2, NextEra Energy Seabrook requested approval of an emergency license amendment that would modify Technical Specification 3.7.4 ACTION b by extending the AOT on a one-time basis for a Service Water system cooling water cell from 7-days to 16-days.

4.2.3 In Reference 6.3, NextEra Energy Seabrook requested withdrawal of Reference 6.2 based Upon subsequent repair efforts on service water cooling tower 51 B fan, a 24-hour endurance run, and declaration of Operability prior to the expiration of the 5-day enforcement discretion period.

4.3 No Significant Hazards Consideration

The proposed license amendment makes an editorial correction to TS Section 1.9. In addition the proposed license amendment modifies Seabrook TS 3.7.4, ACTION b, by eliminating inclusion of one cooling tower cell being inoperable in TS 3.7.4, ACTION b, which results in TS 3.7.4, ACTION b only being applicable for the condition of one cooling tower service water loop being inoperable. The proposed license amendment also adds a new TS 3.7.4, ACTION f, for the condition of one cooling tower cell being inoperable.
The requirements for the condition of one cooling tower cell being inoperable in a new TS 3. 7.4, ACTION fare identical to the requirements for the condition of one cooling tower cell being inoperable in existing TS 3. 7.4, ACTION b, except that the AOT is being increased from 7-days to 21-days.

As required by 10 CFR 50.91(a), NextEra evaluated the proposed changes using the criteria in 10 CFR 50.92 and determined that the changes do not involve a significant hazards consideration. An analysis of the issue of no significant hazards consideration is presented below:

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

Response : No

The proposed change to TS Section 1. 9 is editorial in nature and has no effect on accident scenarios previously evaluated. The proposed change to Seabrook TS 3.7.4, ACTION b results in the AOT being increased from 7-days to 21 -days, which does not affect the redundant, independent cooling tower service water loop or cell such that no single failure will preclude performance of any safety function, and thereby cannot increase the probability of any previously analyzed accident. The proposed changes cannot increase the consequences of any previously evaluated accident since accident analyses assume single failure of the redundant train and the proposed changes do not affect cooling tower cell redundancy. The proposed changes do not affect any accident initiators or Seabrook Station L-2022-176 Docket Nos. 50-443 Enclosure Page 12 of 37

precursors, or alter the design, conditions or configuration of the facility as currently analyzed. All plant equipment will continue to perform consistent with the safety analysis assumptions.

Therefore, the proposed license amendments would not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response : No

The proposed change to TS Section 1.9 is editorial in nature and does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the method governing normal plant operation. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of this adm inistrative change. The proposed change to Seabrook TS 3. 7.4, ACTION b results in the AOT being increased from 7-days to 21-days, which neither modifies plant equipment nor introduces unique operational modes or failure mechanisms. Implementation of the proposed changes do not affect the capability of equipment to perform their respective safety functions. The proposed changes do not alter the types or increase the amounts of fission product effluents predicted in safety analyses and no increase in individual or cumulative occupational exposure will result. No new accident scenarios, transient precursors or limiting single failures will result from the proposed change since all design and performance criteria will continue to be met and the nuclear unit will continue to be operated within the limits of its licensing basis.

Therefore, the proposed license amendments would not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response : No

The proposed change to TS Section 1. 9 is editorial in nature and does npt alter setpoints at which protective actions are initiated or the manner in which safety limits are determined. The proposed change to Seabrook TS 3.7.4, ACTION b results in the AOT being increased from 7-days to 21-days, which does not modify equipment functions, response times or acceptance criteria associated with any accident analyses. No new or altered methods of assessing plant performance are introduced and all accident analysis inputs and assumptions remain unaffected. Thereby, no safety limits or limiting safety settings are challenged by the proposed change.

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

Based upon the above analysis, NextEra 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.
Seabrook Station L-2022-176 Docket Nos. 50-443 Enclosure Page 13 of 37

4.4 Conclusion

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

5.0 ENVIRONMENTAL CONSIDERATION

The proposed license amendment modifies a regulatory requirement with respect to the installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or changes an inspection or surveillance requirement. However, the proposed license amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed license amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed license amendment.

==6.0 REFERENCES==
6.1 NextEra Energy Seabrook, LLC, letter SBK-L -21103, " Request for Exercise of Enforcement Discretion," September 25, 2021 (ADAMS Accession No. ML21268A003)

6.2 NextEra Energy Seabrook, LLC, letter SBK-L -21104, "Emergency License Amendment Request 21-02, Service Water System/ Ultimate Heat Sink One-Time Allowed Outage Time (AOT) Extension, " September 25, 2021 (ADAMS Accession No. ML21268A004)

6.3 NextEra Energy Seabrook, LLC, letter SBK-L-21105, "Withdrawal of Emergency License Amendment Request 21-02, Service Water System/ Ultimate Hea t Sink One-Time
* Allowed Outage Time (AOT) Extension," September 27, 2021 (ADAMS Accession No.
ML21270A107)

6.4 Regulatory Gu ide 1.177, Rev ision 2, Plant-Specific, Risk-Informed Dec ision Making :
Technical Specifications, January 2021 (ADAMS Accession No. ML100910008)

6.5 NRC Policy Statement, "Use of Probabilistic Risk Assessment Methods in Nuclear Activities: Fina l Policy Statement," Fede ral Register, Vol. 60, p. 42622 (60 FR 42622),
August 16, 1995

6.6 Seabrook Station, Unit No. 1 - Issuance of Amendment RE : Administrative Changes to Technical Specifications (TAC No. MB2852), April 3, 2002 (ADAMS Accession No.
ML020510578)

6.7 Florida Power and Light Company, letter L-2019-003, "Application to Revise Technical Specifications to Adopt TSTF 563, "Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program, " March 18, 2019 (ADAMS Accession No.
ML19079A240)

6.8 Point Beach Nuclear Plant, Units 1 and 2; Seabrook Station, Unit No. 1; and Turkey Point Nuclear Generating Station, Unit Nos. 3 and 4 - Issuance of Amendment Nos. 265, 268, Seabrook Station L-2022-176 Docket Nos. 50-443 Enclosure Page 14 of 37

164, 290, and 284 RE: Revise Technical Specifications to Adopt TSTF-563 (EPID L-2019-LLA-0055), February 10, 2020 (ADAMS Accession No. ML19357A195)
Seabrook Station L-2022-176 Docket No. 50-443 Enclosure Page 15 of 37 ATTACHMENT 1

PROPOSED TECHNICAL SPECIFICATION PAGES (MARKUP)

(2 pages follow)
DEFINITIONS

CONTAINMENT INTEGRITY

1.7 CONTAINMENT INTEGRITY shall exist when:
: a. All penetrations required to be closed during accident conditions are either:
: 1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or
: 2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions.
: b. All equipment hatches are closed and sealed, c. Each air lock is in compliance with the requirements of Specification 3.6.1.3,
: d. The containment leakage rates are in accordance with the Containment Leakage Rate Testing Program, and
: e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE

1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATION ~

1.9 CORE ALTERATION shall be the movement ef any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel.
Suspension of CORE AL TERA TIONs shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS REPORT

1.10 The CORE OPERA TING LIMITS REPORT (COLR) provides core operating limits for the current operating reload cycle. The cycle specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.8.1.6. Plant operation within these operating limits is addressed in individual specifications.

DIGITAL CHANNEL OPERATIONAL TEST

1.11 A DIGITAL CHANNEL OPERATIONAL TEST shall consist of exercising the digital computer hardware using data base manipulation and/or injecting simulated process data to verify OPERABILITY of alarm and/or trip functions. The Digital Channel Operational Test definition is only applicable to the Radiation Monitoring Equipment. The DIGITAL CHANNEL OPERATIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.

SEABROOK - UNIT 1 1-2 Amendment No. 48,--84-, 164 PLANT SYSTEMS

3/4.7.4 SERVICE WATER SYSTEM/ULTIMATE HEAT SINK

LIMITING CONDITION FOR OPERATION

3.7.4 The Service Water System shall be OPERABLE with:
: a. An OPERABLE service water pumphouse and two service water loops with one OPERABLE service water pump in each loop,
: b. An OPERABLE mechanical draft cooling tower and two cooling tower service water loops with one OPERABLE cooling tower service water pump in each loop, and
: c. A portable cooling tower makeup system stored in its design operational readiness state.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:
 --- - ---------------- -- ------------- ----- ---- --------- NOTES--------------- ----------- ------------------------
: 1. Enter applicable ACTIONS of LCO 3.8.1.1, "AC Sources - Operating," for diesel generator made inoperable by service water.
: 2. Enter applicable ACTIONS of LCO 3.4.1.3, "Reactor Coolant Loops and Coolant Circulation," for residual heat removal loops made inoperable by service water.
: a. With one service water loop inoperable, return the loop to OPERABLE status within 72 hours, or be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
: b. With one cooling tower service water loop or one cooling tower cell inoperable, return the affected loop or cell to OPERABLE status within+ 21 days, or be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
: c. With two cooling tower service water loops or the mechanical draft cooling tower inoperable, return at least one loop and the mechanical draft cooling tower to OPERABLE status within 72 hours, or be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
: d. With two loops (except as described in c) or the service water pumphouse
, inoperable, return at least one of the affected loops and the service water pumphouse to OPERABLE status within 24 hours, or be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.

SEABROOK - UNIT 1 3/4 7-13 Amendment No. 32, 116, 161 Seabrook Station L-2022-176 Docket Nos. 50-443 Enclosure Page 18 of 37 ATTACHMENT 2

PROPOSED TECHNICAL SPECIFICATION BASES PAGES (MARKUP)

(2 pages follow)
PLANT SYSTEMS BASES

3/4.7.1 TURBINE CYCLE (Continued)

3/4.7. 1.4 SPECIFIC ACTIVITY (Continued)

SURVEILLANCE REQUIREMENTS (SR)

SR 4.7.1.4

This SR verifies that the secondary specific activity is within the limits of the accident analysis. A gamma isotopic analysis of the secondary coolant, which determines DOSE EQUIVALENT 1-131, confirms the validity of the safety analysis assumptions as to the source terms in post accident releases. It also serves to identify and trend any unusual isotopic concentrations that might indicate changes in reactor coolant activity or LEAKAGE.
The surveillance frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES
: 1. 10 CFR 50.67.
: 2. UFSAR, Chapter 15.
..;.-(---------fl 3. SEA-1FJR-2.2-033

3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES

The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blow down in the event of a steam line rupture. This restriction is required to: (1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the Surveillance Requirements are consistent with the assumptions used in the safety analyses.

3/4. 7.1.6 ATMOSPHERIC RELIEF VALVES

The OPERABILITY of the Atmospheric Relief Valves (ARVs) ensures the controlled removal of reactor decay heat during reactor cooldown, plant startup, and after a turbine trip, when the condenser and/or the turbine bypass system are not available. When available, the ARVs can be used to reduce main steam pressure for both hot shutdown and cold shutdown conditions. The ARVs provide a method for cooling the plant to residual heat removal entry conditions should the turbine bypass system to the condenser be unavailable. This is done in conjunction with the Auxiliary Feedwater System providing cooling water from the condensate storage tank (CST).

SEABROOK - UNIT 1 B 3/4 7-8 Amendment No. 92 BC 08 02, 14-05 PLANT SYSTEMS BASES

3/4.7.4 SERVICE WATER SYSTEM/ULTIMATE HEAT SINK (Continued)

Removal of an ocean SW loop from service does not render supported equipment, such as the emergency diesel generator and primary component cooling water system, inoperable. Action a specifies the actions that are required to maintain the plant is a safe condition. [Reference 2]

When a cooling tower pump is operating, interlocks prevent the train associated ocean SW pumps from starting. To provide additional protection during operation on the cooling tower, the ocean SW pump control switches may be maintained in the pull-to-lock position to prevent inadvertent pump operation.
Realigning the SW system to the ocean supply requires manual action, and maintaining the control switches in the pull-to-lock position does not change this required action sequence. Pump operation is not affected by maintaining the control switches in the pull-to-lock position during this period; therefore, OPERABILITY of the SW pumps is not compromised. tw t.,---- -i3 en y-one
: b. With one cooli. g t loop or one cooling to r cell inoperable, Action b provides Se-\\1:ew*days to restore OPERABLE sta s. The sevefl-day allowed outage time is based on a risk evaluation [Reference 4 ], the low Ii ihood of occurrence of a seismic event during the time a cooling tower SW loop or co ling tower cell is inoperable, and operability of the ocean SW system. twenty-one

Removal of the cooling tower SW loop or one cooling tower cell from service does not render supported equipment, such as the emergency diesel generator and primary component cooling water system, inoperable. Action b specifies the actions that are required to maintain the plant is a safe condition. [Reference 2]
: c. Action c provides 72 hours to restore OPERABLE status with two cooling tower SW loops or the cooling tower inoperable. The 72-hour allowed outage time is based on a risk evaluation [Reference 1 ], the low likelihood of occurrence of a seismic event during the time that the cooling tower is inoperable, and operability of the ocean SW system.

Removal of the cooling tower or two cooling tower SW loops from service does not render supported equipment, such as the emergency diesel generator and primary component cooling water system, inoperable. Action c specifies the actions that are required to maintain the plant is a safe condition. [Reference 2]
: d. Action d provides 24 hours to restore OPERABLE status with two ocean SW loops or the SW pump house inoperable. The 24-hour allowed outage time is based on a risk evaluation in Reference 1, the low likelihood of occurrence of a tornado during the time that the ocean SW system is inoperable, and operability of the cooling tower SW system.
Seabrook Station L-2022-176 Docket Nos. 50-443 Enclosure Page 21 of 37 ATTACHM ENT 3

EVALUATION OF RISK IMPACT

1.0 PURPOSE

This document provides a Probabilistic Risk Assessment (PRA) of a proposed license amendment to permanently modify the Seabrook Station (Seabrool<) Technical Specifications (TS) by extending the Completion Time (CT) for an inoperable Service Water Cooling Tower Fan from 7 days to 21 days, which are the subject of Technical Specification Limiting Condition of Operation 3. 7.4.b.

The Service Water System (SW) at Seabrook is designed such that the normal Ultimate Heat Sink (UHS) is the Atlantic Ocean, supplied to the plant through two completely independent and redundant flow trains, each of wh ich consists of two redundant Service Water pumps ( collectively SW-P-41 A, SW-P-418, SW-P-41 C, and SW-P -41 D). The SW system is designed to perform all safety functions during and following all severe natural phenomena, with the potential exception of a seismic event. In the event that seawater flow is restricted due to seismically induced damage to the intake and discharge tunnels, a mechanical draft evaporative cooling tower is provided to dissipate shutdown and accident heat loads.

The cooling tower system provides an alternate source of cooling water which is completely independent of the tunnels and Atlantic Ocean. The cooling tower complex contains a three-cell tower (two cells functional), a five-compartment interconnected basin (three compartments functional), two pump rooms (one room functional) and required valves, piping, and pumps (SW-P-110A and SW-P-1108). It is noted that the non-functional cooling tower substructures are as a result of historically suspended construction of a proposed second unit. The two cooling tower cells contain three fans; two fans are associated with the "B" train (1-SW-FN-518, 2-SW-FN -518) and one fan is associated with the "A" train (SW - FN-51A).

2.0 METHODOLOGY

The FPIE PRA model for this evaluation is documented in SBK-BFJR-22-034, Revision 0, "SBK Application Specific Model for Cool ing Tower License Amendment " [Ref. 7]. This model only includes internal events and was quantified using the default truncation limits demonstrated for convergence.
The internal flooding model of record was obtained from Ref. 2.

Regulatory Gu ide (RG) 1.177, "An Approach for Plant-Specific, Risk Informed Decision-mak ing:
Technical Specifications" [Ref. 1]. RG 1.177 describes acceptable methods for assessing the nature
* and impact of proposed T.S. changes, including AOT extensions, by considering engineering issues and applying risk insights.

Reg. Guide 1.177 directly addresses the PRA-based risk metric requ irements for permanent TS changes, as reproduced below :

 "The staff provides the following TS acceptance guidelines specific to permanent CT changes for evaluating the risk associated with the revised CT, in addition to those acceptance guidelines in RG 1. 174:

a. The licensee has demonstrated that the TS CT change has only a small quantitative impact on plant risk. An ICCDP of less than 1x1& 6 and an ICLERP of less than 1x1& 7 are considered small for a single TS condition entry (Tier 1).
b. The licensee has demonstrated that there are appropriate restrictions on dominant risk-significant configurations associated with the change (Tier 2).
: c. The licensee has implemented a risk-informed plant configuration control program, including procedures to use, maintain, and control such a program (Tier 3)."
Seabrook Station L-2022-176 Docket Nos. 50-443 Enclosure Page 22 of 37 ATTACHMENT 3

EVALUATION OF RISK IMPACT

Based on the available quantitative guidelines for other risk-informed applications, it is judged that the criteria shown in Table 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 (for planned TS entries) 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. This document addresses Tier 1 of the risk associated with the change; Tiers 2 and 3 are addressed in Section 3.4 in the LAR enclosure.

Table 1 - Proposed RislcAcceptance Guidelines

Rislc Acceptance Guideline Basis

ICCDP is an appropriate metric for assessing risk ICCDP < 1E-6 impacts of out of service equipment per RG 1.177.
This guideline is specified in Section 2.4 of RG 1.177.

ICLERP is an appropriate metric for assessing ICLERP < 1 E-7 risk impacts of out of service equipment per RG 1.177. This guideline is specified in Section 2.4 of RG 1.177.

The ICCDP associated with each OOS configuration for a new CT is given by

ICCDPFNS1= (CDFFN51 - CDFBASE) X CT NEW [Eq. 1]
Where
* CDFFNs1= the annual average CDF calculated for configuration with equipment OOS (all quantified hazards)
* CDFsAsE = baseline annual average CDF with average unavailability for all equipment. This is the CDF result of the baseline PRA (all quantified hazards)
* CT NEW= the new extended CT (in units of years)

NOTE: ICCDP is a dimensionless probability and ICLERP is quantified similarly.

3.0 PRA QUALITY

The ASME / ANS PRA Standard (ASME/ANS RA-Sa-2009), Ref. 8, has technical elements, high-level requirements (HLRs), and detailed supporting requirements (SRs). NRC Regulatory Guide 1.200 Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities" [Ref. 5], endorses the standard with minor clarifications. The EPRI ePSA database includes each supporting requirement from the standard along with the clarifications from RG 1.200.
Seabrook Station L-2022-176 Docket Nos. 50-443 Enclosure Page 23 of 37 ATTACHMENT 3

EVALUATION OF RISK IMPACT

The Seabrook PRA has undergone peer review against the ASME PRA Standard, Parts 1 (configuration control), 2 (internal events), and 3 (internal flood events).

Peer reviews have been conducted against internal event supporting requirements as follows:
In 1999, a review of all technical elements was performed using the industry PSA Certification process, the precursor to the PRA Standard.
* In 2005, a focused peer review was performed for the elements AS, SC, and HR, as well as configuration control. This review was done to PRA Standard ASME RA-Sa-2003.

In 2009, a focused peer review was performed for all elements of Part 3, Internal Flooding. This review was done to PRA Standard ASME/ANS RA-Sa-2009.

In 2012, a focused peer review was performed for the element LE. This review was done to PRA Standard ASME/ANS RA-Sa-2009.

In 2019, a focused peer review was performed on all elements upgraded by the conversion from Riskman to CAFTA. This review was done to PRA Standard ASME/ANS RA-Sa-2009.

Four self-assessments against the internal event SRs in the PRA stand.ard were performed in 2005 (ASME RA-Sa-2003), 2007 (ASME RA-Sb-2005), 2010 (ASME/ANS RA-Sa-2009) and 2011 (ASME/ANS RA-Sa-2009). The first three self-assessments considered all internal events technical elements. The SA-2011 addressed only the open findings against specific SRs.

The 2011 self-assessment represents the most current status of Seabrook PRA capability, except for element LE. The 201 O self-assessment had assessed the 2009 PRA against each of the 254 internal events supporting requirements in ASME/ANS RA-Sa-2009. That assessment reviewed the results of previous peer reviews and their observations along with the subsequent revisions to the PRA that addressed the observations.

In October 2017, all resolved findings were reviewed to Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13, "Close-out of Facts and Observations" (F&Os) as accepted by the NRC staff in their May 3, 2017, memorandum (ML17079A427).

Following the findings closure review, the 2019 focused scope peer review identified additional findings.
Appendix B provides a listing of the remaining open findings and the status of their resolution as well as an assessment of the impact on this evaluation. Overall, the open findings do not have an impact.

Finally, the open Items in the Seabrook Change Database [Ref. 4] were examined for their potential impact on the risk analysis for this LAR. The open items involved non-significant enhancements or corrections to the model or documentation changes. The documentation updates have no impact. The enhancement and correction items were reviewed and judged to have minimal impact on the risk analysis of this LAR.

4.0 ASSUMPTIONS
: 1. Maintenance on redundant Service Water trains would not be occurring. Performing maintenance on concurrent trains would cause entry into more restrictive actions of LCO 3.7.4 (for example, having one cooling tower train and one ocean cooling train would cause entry into 3.7.4.d, which Seabrook Station L-2022-176 Docket Nos. 50-443 Enclosure Page 24 of 37 ATTACHMENT 3

EVALUATION OF RISK IMPACT

requires restoration within 24 hours). A restriction on voluntary redundant maintenance is a required Risk Management Action (RMA).
: 2. In the event maintenance is occurring on a Service Water train when a cooling tower train randomly fails, entry would also be made into a more restrictive action of LCO 3.7.4.
: 3. While the LAR also restricts concurrent maintenance on the Emergency Diesel Generators (EDG) and RPS, ESFAS, and containment isolation instrumentation, these compensatory measures are not quantitatively credited in this assessment.

5.0 COMMON CAUSE EVALUATION

This evaluation does not examine any specific failure mode or ongoing issue of either cooling tower fan; in addition, the risk impact of each cooling tower fan out of service is assessed. As such, the existing CCF probabilities are most appropriate to use in this evaluation and no CCF penalty is warranted.

6.0 EVALUATION

The three service water cooling tower fans are individually modeled in the Seabrook PRA. Each fan has associated basic events of fail to run, fail to start, and fail to run when the train is normally operating. Loss of a fan fails the associated cooling tower train, and subsequently is one of two inputs into an AND gate for failure of one train of the Service Water system; that is, failure of the cooling tower service water train AND Atlantic Ocean service water train results in a failure of that service water train. In the event of a failure of a cooling tower fan only, the associated normal (Atlantic Ocean) service water train would be available; if that service water train additionally failed, then the entire second service water train would remain available (with both the second normal train and second cooling tower train available). It is noted that the cooling tower pumps (11 0A and 11 OB) have an equivalent impact on the model to the fans, such that the fans reasonably represent a "train" of cooling tower service water. The high degree of redundancy leads to the expectation that the risk significance of the cooling tower fans individually will be low.

As discussed in Assumptions 4.1 and 4.2, maintenance events on alternate cooling train components were restricted (i.e., maintenance events were set to FALSE for quantification). Appendix A contains the flag file modifying this quantification from the model documented in Ref. 7, including an explanation of maintenance events that were set to FALSE. Because (from a quantitative-only standpoint) not performing maintenance on the service water pumps may provide a benefit to the overall CDF, the "baseline" model was run with the SWS maintenance events false to ensure the change in risk was accurately calculated.
Seabrook Station L-2022-176 Docket Nos. 50-443 Enclosure Page 25 of 37 ATTACHMENT 3

EVALUATION OF RISK IMPACT

7.0 RESULTS

7.1 Quantification

The internal events PRA quantification was performed on the Ref. 7 FPIE model with a CDF truncation of 1 E-12 and LERF truncation of 1 E-13. Table 2 shows the results of the FPIE model quantification.

Table 2 - Results for FPIE Delta Risk Calculation

Case CDF(/yr) Delta CDF LERF{/yr) Delta LERF
(/yr) (/yr)

Baseline 2.40E-06 1.0SE-07

SW-FN-51A failed 2.42E-06 1.4E-08 1.0SE-07 3E-11

1-SW-FN-518 failed 2.41E-06 8.2E-09 1.0SE-07 8E-12

2-SW - FN-518 failed 2.41E-06 8.2E-09 1.0SE- 07 8E-12

Table 3 provides the results of the Internal Flooding model utilizing the 2019 model update [Ref.
2]. It is noted that the slightly different, albeit negligible, change in risk between trains is due to the alignment of the SEPS to the B train, causing a slight increase in risk for the B SW train due to mitigation abilities when the A train is lost.

Table 3 - Results for Internal Flooding Delta Risk Calculation

Case CDF{/yr) Delta CDF LERF(/yr) Delta LERF
(/yr) (/yr)

Baseline 1.07E- 06 2.63E-10

SW - FN-51A failed 2.51E- 06 1.4E-06 2.99E-10 4E-11

1-SW-FN-518 failed 2.S0E- 06 1.4E-06 2.99E-10 4E-11

2-SW - FN-51B failed 2.S0E-06 1.4E-06 2.99E-10 4E-11 Seabrook Station L-2022-176 Docl<et Nos. 50-443 Enclosure Page 26 of 37 ATTACHMENT 3

EVALUATION OF RISK IMPACT

The contribution to change in risk for each individual fan out of service was then summed to determine the total contribution from internal events and flooding in Table 4. While nearly identical, the delta CDF from fan 51A was the bounding condition.

Table 4 - Conditional Risk Calculation for Internal Events and Internal Flooding

PRA Metric CTnew {yrs) Delta CD F {LERF) 21 DayICCDP
(/yr) (IE+ IF) (ICLERP)

CDF 0.058 {21 days) 1.45E-06 8.3E-08

LERF 0.058 (21 days) 6.8E-11 3.9E-12

The internal events contribution of a single cooling tower fan out of service is relatively small.

7.2 Importance Measures

Because support systems and other modeled events could underpin these events, model importance measures were further reviewed. The ratio of configuration-specific Risk Achievement Worth (RAW) and baseline model RAW was considered a noteworthy figure. If a RAW value increases significantly between the baseline and configuration-specific models, it would be potentially noteworthy in terms of plant configuration control. The importance measures from the FN-51A failure case were utilized (due to the slightly higher overall increase in risk when including Internal Flooding); it is noted that due to train symmetries that the discussion below is applicable to the opposite train components in the event of a FN-51 B failure.

Table 5 shows the ratio of configuration-specific Risk Achievement Worth (RAW) to baseline model RAW (RAWFN51/RAWBase) for the five CDF events with the highest ratio.

As expected, the top change in event importance are valve failures that would fail the B train of service water; the two valves SW-V-65 and SW-V-67 would fail both the ocean cooling and cooling tower trains, resulting in a loss of all service water concurrent with the A CT out of service and failure to transfer to the A cooling train. This leads into the next two important operator action combinations; combinations 247 and 201 involve transferring between the ocean cooling and cooling tower trains of service water, joined with failure to trip the Reactor Coolant Pumps (resulting in an RCP seal LOCA).

When reviewing LERF importance measures, no additional components were identified with an unexpected increase in importance measures.

The importance measures for the internal flooding events were also investigated, due to the slightly higher change in risk. The top event increasing in importance for the internal flooding evaluation were all related - floods due to rupture of the opposite service water train (which fails that train of service water); and human error combinations related to tripping the reactor coolant pumps (RCPs) given a loss of seal cooling water, preventing seal LOCAs.
Seabrook Station L-2022-176 Docket Nos. 60-443 Enclosure Page 27 of 37 ATTACHMENT 3

EVALUATION OF RISK IMPACT

Fussell-Vesely (FV) values were reviewed for internal events and floods. For both internal events and internal flooding, the primary increase in importance involved restoring RCP seal cooling and tripping the RCP pumps to prevent seal LOCAs.

In summary, the review of importance values demonstrated that, as expected, the opposing trains of service water become more important in the event a cooling tower fan is out of service.
However, outside of restricting maintenance on operating trains, no additional compensatory actions were identified from the importance measure review.

Table 5 - Comparison of Importance Measures

Basic Event BE Description CDFRAW Ratio

SWV65.TCNO STRAINER Train B Inlet SWV65 transfers closed 12.3

SWV67.TCNO STRAINER Train B Outlet SWV67 transfers closed 12.3

COMBINATION_247 Dependent HEP for 7.3 H H.SWOCCT. FA,H H.OTRCPlZ. FA,H H.OAL Tl.FL

COMBINATION_201 Dependent HEP for 6.8 HH.SWOCCT.FA,HH.OTRCPl.FA,HH.OALTl. FL

WWW-XFRA Transfer to CT During Train A Surveillance 3.8

7.3 Configuration Management

In terms of risk significance of plant configurations, RG 1.177 requires reasonable assurance that risk-significant plant equipment outage configurations will not occur when specific plant equipment is out of service consistent with the proposed change.
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As it is recognized that plant configuration (maintenance and typical train lineups) involving service water are most important to this completion time assessment, a sensitivity evaluation considering average maintenance on non-failed service water trains was performed. The importance measures for this configuration were then reviewed similar to the above cases.
Because the overall change in LERF was insignificant, this sensitivity is not performed for LERF.
It is expected the change in LERF would correlate relative to the change in CDF.

Table 6 - Results for Sensitivity Delta Risk Calculation - No Maintenance Restrictions on Service Water Components_

Case IECDF{/yr) Delta CDF (/yr) IFCDF Delta CDF (/yr)

Baseline 2.4SE-06 1.9SE-06

FN -51A Failed without Maint. Restrictions on Remaining SW Components 8.06E-06 S.61E-06 6.41E-06 4.46E-06

As would be expected, the change in CDF is more significant when train maintenance is not restricted. A review of importance measures for this configuration demonstrates that the most risk significant components and line-ups are the redundant service water trains. This is an important sensitivity as it demonstrates that if and when the LCO is intentionally entered for fan maintenance, then maintenance and equipment line-ups should be adjusted accordingly by restricting maintenance on the remaining service water and cooling tower components.

8.0 EXTERNAL EVENTS

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 risk assessment. Internal events, including internal flooding, are quantitatively addressed as described in the previous sections. The impact due to seismic, high winds, external floods, shutdown operation, and other hazard groups are addressed here.

Seabrook completed an Individual Plant Examination of External Events (IPEEE) in 1992 [Ref.
1 O]. Section 1.5 of the I PEEE summarizes the major findings and states that fire and seismic events were the only important contributors to external events core damage. The CDF contribution due to fire was calculat.ed to be 1.2E-05 per year. Since Fire PRA methodology has evolved significantly, the qualitative IPEEE assessment is not readily utilized. Therefore, a conservatively bounding risk assessment of fire, using surrogate events In the FPIE model, is performed.
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The bounding calculation determines the fire initiating event frequency, utilizes an appropriate gate for event mitigation, and then calculates the impact on risk metrics.

The fire ignition frequencies at Seabrook have been calculated from NUREG-2169. The frequencies for all ignition sources, except for Bin 20 (Off-Gas/Hydrogen Recombiner at BWR),
were summed resulting in a total site fire frequency of 2.0E-01 used for fires in plant areas except rooms directly affecting the ocean cooling water system (Fire Ignition Frequency, FIF). Fires in rooms containing the cooling tower equipment are not evaluated as they would cause a loss of the system, such that the delta risk would be negligible. No credit for severity factors or non suppression probabilities is applied.

The gate selected for a bounding assessment is SWAB - Failure of Both SW Trains. The average failure probability of SWAB was calculated. Then, various combinations of fire initiating events based on the Appendix R analysis [Ref. 3] were analyzed. Specifically, three scenarios/ delta risks were calculated and summed together. First, fire events in the plant that did not specifically affect the Service Water system. Based on a review of the Appendix R report, this would constitute fires in 87 out of 91 fire areas. Second, fire events that fail a single train of SW were calculated; fires in 3 out of 91 fire areas cause this condition (SW-F-1 B-A, SW-F-1 C-A, and SW F-1 D-A in the App. R report). In the Appendix R report, a fire in these fire areas fail both trains of Service Water. However, this failure occurs due to a failure of SW room ventilation. As documented in the Seabrook Systems Analysis Notebook Section 10.2.1, failure of SW room ventilation does not cause a significant temperature increase or SW equipment failure. SW-F-1 B A and SW-F-1 C-A each fail one train of SW pumps due to direct cable impact; it is noted that SW F-1 D-A does not fail either train of SW directly and could be included with the bulk plant fire areas but is conservatively assumed to fail one train. Finally, scenario SW-F-1 E-Z could fail both trains of Service Water; this is a single scenario out of the 91.

For the four SW events of interest, the FIF was calculated as (values from NUREG/CR-2169, Table 4-6, total for plant) Bin 14, electric motors (5.43E-03) + Bin 15, electrical cabinets (3.00E-
: 02) + Bin 21, pumps (2.72E-02) + Bin 23, transformers (9.56E-03) + Bin 26, ventilation subsystems (1.64E-02) applying a conservative fraction of the frequency for these bins for the number of components in the zone over the total number in the plant; a factor of 4/100 for pumps, motors, ventilation units and transformers and a factor of 4/500 for electrical panels; results in a total Fixed Ignition Frequency bounding value of 2.58E-03. The transient frequency is based on the sum of Bin 3, control aux building (3.33E-03) + Bin 25, plant wide (8.54E-03) + Bin 37 (6. 71 E-03), divided by 91 (total number of plant zones, results in a transient frequency of 2.04E-04. This is conservative given that the transient loading in the SW fire zones is expected to be significantly less than the average plant fire zone transient frequency. The bounding frequency for the SW pump zones is 2.78E-03. This value is multiplied by 3 for the single-train SW failure events, as a fire in 3 of the zones fails one train of service water. For fire zone SW-F-1 E-Z, it is assumed that half of the fires would fail both trains of service water, while half would fail one train of service water; this is a reasonable assumption given the unlikelihood of the magnitude of a fire required to impact multiple trains.

The change in risk was calculated for both fans ; SW-FN-51A is reported below as it had the higher contribution. The delta CDF contribution from these fire scenarios was calculated as follows:
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h.CDF = FIF
* l!.ProbswAB Table 7-Results for Bounding Fire PRA Assessment

FN-51A Case FIF Base Probability Failed iJ.Prob iJ.CDF ICCDP at 21 Days Prob.

Fire in a plant area not affecting 0.206 7.27E-07 7.48E-07 2.lE-08 4.3E-09 2.SE-10 Service Water (87 fire zones), total for all bins from NUREG/CR-2169

Fire failing one train of Service Water 9.73E-03 1.37E-04 1.40E-04 3.8E-06 3.7E-08 2.lE-09 (3 fire zones+ 1/2 SW-F-1E-Z)

Fire failing both trains of Service 1.40E-03 3.16E-03 9.48E-03 6.3E-03 8.9E-06 S.lE-07 Water (1 fire zone)

Due to the low LERF results in Internal Events and Internal Floods, the contribution to LERF from fire is considered negligible.

Seismic

Seabrool< does not have a seismic PRA. However, the increase in risk during a seismic event is deemed negligible. The station essentially has four redundant trains of service water; while the ocean intake tunnel has not been seismically qualified, the Safety Analysis Report states thaUhe tunnel would require more than 95% blockage prior to changeover to the cooling towers would be required. The cooling tower systems themselves are seismically qualified and redundant fans would operate. Given the typically low initiating event frequency of earthquakes greater than the safe shutdown earthquake, the change in seismic risk with one cooling tower fan out of service would be insignificant.

High Winds

Seabrook does not have a high winds PRA. However, the cooling tower systems are designed and qualified for tornado loads, and the ocean-supplied service water pumps are located in a structure designed to protect against wind and tornado loading. As such, there is a high degree of confidence that the increase in high winds risk is negligible as a result of this change.

Shutdown Operations

During shutdown operations, ris_k is managed in accordance with station procedures.
Qualitatively, the absence of one cooling tower fan would be less risk significant than during normal operation, as the heat load to remove would be less and, in general, the time available to restore alternate cooling sources is longer. The change in shutdown risk is negligible.
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Other External Events

Seabrook does not explicitly evaluate any other external event hazards (e.g., external flooding) in the probabilistic modeling. The potential impact associated with other external events is judged to be bounded by the evaluations in this document.

9.0 

==SUMMARY==
AND RECOMMENDATIONS

Extension of the service water cooling tower fan allowed outage time of 7 days to 21 days has a minor impact on the PRA results and the increase is within NRC limits established by Reg. Guide 1.177.

Table 8 - Conditional Risk Calculation

Incremental Probability for 21 Day AOT PRAMetric
.................................................... ICCDP ICLERP

....Internal Events 1.4E-08 1.9E-12................................................

........................Internal Floods 8.3E-08 2.0E-12............................

.................Fire 5.1E-07 Neg..........................,.........

External Events Negligible Neg.

Total 5.93E-07 3.9E-12 Acceptance Criteria 1.00E-06 1.00E-07

Based on the model insights and sensitivities, the compensatory measures (i.e., Risk Management Actions - RMAs) for operational consideration :

 + In the event of a cooling tower fan failure, immediately suspend any maintenance on service water and cooling tower components, restoring to service as soon as possible,
 + Prior to removing a cooling tower fan from service, verify Operability of the redundant cooling tower train, and
 + Prior to removing a cooling tower fan from service, verify Operability and alignment of the ocean-supplied service water trains.

==10.0 REFERENCES==
: 1. USNRC, Regulatory Guide (RG) 1.177, "An Approach for Plant-Specific, Ri.sk Informed Decision making: Technical Specifications," Revision 2, (ADAMS Accession No. ML100910008).
: 2. Report 027144-RPT-02, "Seabrook Nuclear Station Quantification Results," Revision 0, September 2019.
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: 3. Report "Fire Protection of Safe Shutdown Capability (1 0CFR50, Appendix R), Revision 18.

4. Seabrook Model Change Database "SEAModel Change Database Updated for SBK20 and future.accdb," reviewed Sept. 2022
: 5. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities," Revision 2, March 2009.
: 6. Seabrook Station, Unit No. 1 - Issuance of Amendment Regarding the Risk-Informed Justifications for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (TAC No. MF1958) (ADAMS Accession No. ML13212A069)

7. Report SBK-1FJR 042, "Seabrook Internal Events PRA Model Update," Revision 1, April 2022.
: 8. ASME PRA Standard RA-Sa-2009, Addenda to ASME/ANS RA-S-2009 Standard for Level 1 /
Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, February 2009. *
: 9. NextEra Licensing Document L-2022-176, "License Amendment Request to Revise Cooling Tower Cell Requirements."
: 10. North Atlantic Energy Service Corp. (NAESC), 1992, "Individual Plant Examination External Events Report for Seabrook Station," Octob_er 2, 1992 (ADAMS Accession No : ML080100029).

11.0 APPENDICES

APPENDIX A MODEL QUANTIFICATION SETTINGS APPENDIX 8 SEABROOK OPEN PEER REVIEW ISSUES Seabrook Station L-2022-176 Docket Nos. 50-443 Enclosure Page 33 of 37 ATTACHMENT 3

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SW-FN-51A OUT OF SERVICE (OOS) FLAG
 ; Maintenance of SWS Pumps would be limited by TS 3.7.4.d if CT FN OOS WWW-MJM41A EQU.F.
WWW-MJM418 EQU.F.
WWW-MJM41C EQU.F.
WWW-MJM41D EQU.F.
WWW-MN41A EQU.F.
WWW-MN418 EQU.F.
WWW-MN41 C EQU.F.
WWW-MN41D EQU.F.

 ; Maint of alt-train CT pump would be limited by TS 3. 7.4.b if CT FN OOS WWW-MN1108 EQU.F.

 ; Normal alignments of P-110A in-service would not be utilized if FN-51A OOS WWW-NORMEB EQU.F.
WWW-NORMED EQU.F.
WWW-NORMEF EQU.F.

 ; Take FN-51A OOS for all scenarios SWFN51A.FR EQU.T.
SWFN51A.FRNO EQU.T.
SWFN51A.FRNO.YR EQU.T.
SWFN51A.FS EQU.T.

SW-FN-518 OOS FLAG
 ; Maintenance of SWS Pumps would be limited by TS 3.7.4.d if CT FN OOS WWW-MJM41A EQU.F.
WWW-MJM41B EQU.F.
WWW-MJM41C EQU.F.
WWW-MJM41 D EQU.F.
WWW-MN41A EQU.F.
WWW-MN418 EQU.F.
WWW-MN41C EQU.F.
WWW-MN41D EQU.F.

 ; Maint on alt-train CT pump would be limited by TS 3.7.4.b if CT FN OOS WWW-MN110A EQU.F.

 ; Normal alignments of P-1108 in-service would not be utilized if FN-518 OOS WWW-NORMAF
* EQU.F.
WWW-NORMCF EQU.F.
WWW-NORMEF EQU.F.

 ; Take FN-518 OOS for all scenarios SWFN51 B.FR EQU.T.
SWFN518.FRNO EQU.T.
SWFN51 B.FRNO.YR EQU.T.
SWFN51 B.FS EQU.T.
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2-SW-FN-51 B OOS FLAG
 ; Maintenance of SWS Pumps would be limited by TS 3.7.4.d if CT FN OOS WWW-MJM41A EQU.F.
WWW-MJM41B EQU.F.
WWW-MJM41C EQU.F.
WWW-MJM41D EQU.F.
WWW-MN41A EQU.F.
WWW-MN41B EQU.F.
WWW-MN41C EQU.F.
WWW-MN41D EQU.F.

 ; Maint on alt-train CT pump would be limited by TS 3.7.4.b if CT FN OOS WWW-MN110A EQU.F.

 ; Normal alignments of p.:11 OB in-service would not be utilized if FN-51 B OOS WWW-NORMAF EQU.F.
WWW-NORMCF EQU.F.
WWW-NORMEF EQU.F.

 ; Take 2FN-51 BOOS for all scenarios 2SWFN51B.FR EQU.T.
2SWFN51 B.FRNO EQU.T.
2SWFN51B.FS EQU.T.
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Appendix B Seabrook Open Peer Review Issues

Supporting Issue Evaluation Requirement

The Seabrook PRA uses all operating experience when performing the Bayesian update. The use of all operating experience in the Bayesian update Section 1.0 of the Seabrook can provide non-conservative results for component failure probabilities. For PRA Notebook 13, Data DA-D4-01 example, if a component has been replaced, previous operating experience is Analysis, was revised (to Rev. 2)
DA-E2-02 no longer applicable for that component. to provide additional information.
The Bayesian reasonableness check does not discuss any criteria for when No impact on application.

there are 0 failures in the plant-specific experience. For these cases, none of the checks will pass the specified criteria.

The following documentation issues were identified:
: 1) Table 13.6-1 of the Data Analysis shows the Bayesian validation of the Seabrook type codes. It is noted that the Bayesian update equations used for Beta distributions are incorrect. The equation used to update the beta parameter of the beta distribution should be B_prior + n_exposures -
n_failures. The current equation used is B_prior + n_exposures. Note that the The Data Notebook was revised DA-E1-01 current equation used is not consistent with the CAFTA Bayesian update tool. to address this documentation
: 2) Section 13.6.2 of the Data Analysis discusses three conditions for only finding. No impact on checking the reasonableness of the Bayesian update. In the description of application.
the conditions ii should be stated '... 5th percentile and less than the 95th percentile of the generic/posterior distribution.'
: 3) Section 13.6.2 states that the parameters of Interest in the reasonableness check are the: mean values, 5th percentile value, and 95th percentile value.
Table 13.6-1 does not provide the mean values.
The following documentation issues were identified:
: 1) A review of the CAFTA database shows that there are 6 common cause groups making use of the MGL method : BUSFX, BUSFL, LINES, LINES.YR, LINESMNT, and LINESMNT.YR. A search of the System Analysis notebook states that for B.US56FX 'Note that MGL CCF parameters a re used in the The Data Notebook was revised DA-E2-01 2019 update because the 2015 update to NUREG/CR-5497 did not have to address this documentation information on switchgear CCF failure data.' This statement does not provide only issue. No impact on a reference to the data source used, and the data notebook does not provide application.
this information either.
: 2) There is no discussion regarding the selection of staggered or non staggered testing schemes and the use of these calculation methods for the CCF groups.
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The Human Reliability Analysis 1)Self-assessment identifies limitations with manpower requirements and (HRA) was revised in support of there still appear to be gaps w ith HRAC specific inputs for manpower. the 2022 Seabrook model Additionally, execution locations are also not identified fo r all actions. update. The three items 2)Not being able to reproduce results. Recreating the dependency analysis identified were all addressed in using the same cutsets that were used and creating a combination event Revision 2 of SBK-1FJR-19-041, recovery rule file resulted in 860 combinations versus the documented 505 "Seabrook Human Reliability QU-C2-01 / HR-G?- combinations in the Section 11 HR document Section 11.8.1.3. Analysis - Internal Events."
01 3)Manual combination and dependency overrides lacked sufficient Execution locations were added justification for assigned dependency levels. For example, combination of to the HRA Calculator, the HH.OFL0CW.FL and HH.OFL 1 CW.FA, the cur rent justification taken is for method for performing larger timing separation between actions, however, the override taken is dependency analysis was equiva lent to intervening success. This isn't sufficient justification for the documented and reproducible, override taken. and dependency overrides were justified. No impact on application.

Section 3.0 of Section 11, Human Actions Analys is, discusses methodology This issue is documentation and references PRA-106 "PRA Model Guidelines", Section 106E only. The methods used for AS-C1-01 / DA-E1-Methodology for Human Reliability Analysis. PRA-106 is the modeling performing HRA and 02 / HR-I1-01 / SY information for RISKMAN. No discussion could be found for dependency dependency analysis are C1-01 analysis methodology in the co n version report. documented in Rev. 2 of SBK-Similar issue was found to exist in Systems Analysis, Data Analys is, HRA, 1 FJR-19-041. No impact on and Accident Sequence. application.

There are instances where the information from Appendix 11.1A does not This issue was resolved with match the HRAC. See example below. Rev. 2 of SBK-1 FJR-19-041.
HH.OHSB1.FA Tcog 5 minutes versus Appendix 11.A 1 Tcog of 20 -30 Operator interviews were re HR -E4-01 minutes. performed in support of the Also, Operator interview Insights in HRAC for HH.OAL T1.FL don 't seem to report revision and updated in match the interview documentation. This appears to be a systemic problem the HRA Ca lculator. No impact as there were other lnstances *found. on application.

Logic flags have not been set to TRUE or FALSE for all flags prior to the This issue was resolved by flag generation of cutsets. The current methodology sets logic flags to TRUE in file changes in the 2022 model QU-B9-01 the recovery rules which occurs after the generation of cutsets. Additional update. No non-minimal cutsets cutsets have been generated in the final results that should not exist as they were identified in the model are non -minimal. update. No impact on application.

SOKC is not accounted for in some type codes that use identical data sets. Th is issue was resolved w ith the One example Is for the type codes NICB1 C and NICB1O. Both of these type 2022 model update. Type codes QU -E3-01 / QU-A3-codes use the same data set, but since they are different type codes that shared data sets were 01 UNCERT does not take the same sample for both distributions. This appears revised to ensure functionality to be a common approach when the generic data doesn't delineate between with UNCERT. No impact on the different failure modes of a component. application.
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: 1) The FTREX.lnl file was not documented. This is necessary to quantify the model and is significant because the default method is not used.
: 2) The criteria establishing convergence is based on <=5% change when compared to the next decade. The example in the standard uses a <5% final change. The final change Is interpreted as calculating the percent change at the current truncation level with respect to the previous decade truncation level not the next. The criteria used is adequate, but there is no QU-F2-01 documentation of definition used to establish convergence. This issue is documentation 3)There is no discussion of the lop basic events and why they make logical only. No impact on application.
sense. A general statement that notes that basic events importance's were reviewed to ensure they make logical sense is not sufficient evidence for the actual review taking place.
: 4) There is no documentation of how the circular logic is broken. A demonstration was performed that identified a couple examples of where in the model circular logic was broken. This identification and modeling technique needs to be documented.

This issue was resolved w ith the Component importance measures were not identified. The supporting 2022 model update. Component QU-D6-01 requirement specifically requires the identification of significant SSCs. Importance measures were identified in the model update report. No impact on application.}}