Text

Enclosure 1 - List of Revised Required Actions to Corresponding PRA Functions and Additional Supporting Information
	ML22140A142
Person / Time
Site:	Point Beach
Issue date:	05/20/2022
From:	; Point Beach
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML22140A131	List: ML22140A132; ML22140A133; ML22140A134; ML22140A135; ML22140A136; ML22140A137; ML22140A138; ML22140A139; ML22140A140; ML22140A141; ML22140A142; ML22140A143; ML22140A144; NRC 2022-0007, Enclosure 11 - Monitoring Program;
References
	NRC 2022-0007
	Download: ML22140A142 (38)
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Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Page 1 of 27 ENCLOSURE 1 Point Beach Units 1 and 2 License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b List of Revised Required Actions to Corresponding PRA Functions and Additional Supporting Information

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 Page 2 of 27

1.

INTRODUCTION In accordance with the NRCs revised model application for TSTF 505, Revision 2 (Reference 1), Enclosure 1 of the amendment request is to provide confirmation of the PRA models, including the necessary scope of structures, systems, and components (SSC) and their functions for each proposed application of the Risk-Informed Completion Time (RICT) Program to the TS LCO Conditions and Required Actions.

This enclosure provides confirmation that the Point Beach PRA model includes the necessary scope of SSCs and their functions to address each proposed application of the RICT Program. The enclosure addresses the applicable design basis functions, how each are modeled in the Point Beach PRA, and provides justification for the use of proposed surrogates to adequately capture configuration risk, where applicable. The enclosure also provides a comparison of the success criteria used in the PRA model to the design basis success criteria at a train and component/parameter level. The comparison addresses each of the TS LCO Conditions and associated Required Actions proposed for the Point Beach RICT Program, as identified in the TS markup pages of Attachment 2 of this amendment request. Also provided are additional justifications for the specific TS Required Actions recommended in the NRCs Final Revised Model Safety Evaluation for TSTF-505, Revision 2 (Reference 5) and information to support instrumentation redundancy and diversity, as also recommended in the NRCs Revised Model Safety Evaluation.

1.0 SCOPE Table E1-1 below lists each TS Required Action Condition proposed for the Point Beach RICT Program and documents information regarding the associated SSCs credited in plant safety analyses, the analogous PRA functions, and the results of the comparison. The Comments column provides where applicable, a disposition of inconsistencies between the TS and PRA functions regarding the scope of SSCs and the success criteria. Where the PRA scope of SSCs is not consistent with the TS, additional information is provided to describe how the Required Action Condition can be evaluated using appropriate surrogate events which permit a risk evaluation to be completed using the Configuration Risk Management Program (CRMP) tool for the RICT Program. Differences in success criteria typically arise due to the RG 1.200 (Reference 2) requirement to employ realistic as-built, as-operated criteria, whereas design basis criteria are necessarily conservative and bounding. These differences are addressed to demonstrate that the PRA criteria provide a realistic estimate of the risk of the TS condition as required by NEI 06-09 (Reference 3).

References:

1. NRC Revised TSTF 505, Revision 2, Model Application, (ADAMS Accession No. ML18115A482)

2. Regulatory Guide (RG) 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, March 2009 (ADAMS Accession No. ML090410014)

3. NEI 06-09 (Revision 0), Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Industry Guidance Document (ADAMS Accession No. ML063390639)

4. NRC letter to NEI, Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, Risk-Informed Technical Specifications Initiative 4B, Risk-Managed Technical Specifications (RMTS) Guidelines (TAC No. MD4995), dated May 17, 2007 (ADAMS Accession No. ML071200238)

5. NRC Safety Evaluation, Final Revised Model Safety Evaluation of Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, dated November 21, 2018 (ADAMS Accession No. ML18269A041)

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 Page 3 of 27 Table E1 List of Revised Required Actions to Corresponding PRA Functions Required Action Condition Description Applicable SSCs SSC Function(s)

PRA Modeled Design Success Criteria PRA Success Criteria Comments TS 3.3.1, RPS Instrumentation (Table 3.3-1)

Condition B One Manual Reactor Trip channel inoperable (Modes 1,2)

(FU1)

Two Manual Rx Trip channels Reactor Trip Initiation No 1 of 2 Manual Rx Trip channels Not Modeled - See comments The operator actions for failure to actuate a manual reactor trip will be used as a surrogate to conservatively bound the risk increase associated with this function as permitted by NEI 06-09.

Condition D One channel inoperable (FU2a)

Four Power Range Neutron Flux High channels Reactor Trip Initiation No 2 of 4 Power Range Neutron Flux-High channels Not Modeled - See comments The condition of one of two inoperable reactor trip breakers is used as a surrogate for this condition.

(FU2b)

Four Power Range Neutron Flux Low channels Reactor Trip Initiation No 2 of 4 Power Range Neutron Flux-Low channels Not Modeled - See comments The condition of one of two inoperable reactor trip breakers is used as a surrogate for this condition.

(FU5)

Four Overtemperature T channels Reactor Trip Initiation No 2 of 4 Overtemperature T channels Not Modeled - See comments The condition of one of two inoperable reactor trip breakers is used as a surrogate for this condition.

(FU6)

Four Overpower T channels Reactor Trip Initiation No 2 of 4 Overpower T channels Not Modeled - See comments The condition of one of two inoperable reactor trip breakers is used as a surrogate for this condition.

(FU7b)

Three Pressurizer Pressure -

High channels Reactor Trip Initiation No 2 of 3 Pressurizer Pressure - High channels Not Modeled - See comments The condition of one of two inoperable reactor trip breakers is used as a surrogate for this condition.

(FU13)

Three SG Water Level Low-Low channels per SG Reactor Trip Initiation No 2 of 3 SG Water Level Low-Low channels on any SG Not Modeled - See comments The condition of one of two inoperable reactor trip breakers is used as a surrogate for this condition.

(FU14)

Two SG Water Level Low; Coincident w/ Steam-Flow/ Feed-Flow Mismatch channels per SG Reactor Trip Initiation No One SG Water Level Low coincident w/

one Steam-Flow/

Feed-Flow Mismatch channel on any SG Not Modeled - See comments The condition of one of two inoperable reactor trip breakers is used as a surrogate for this condition.

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 Page 4 of 27 Table E1 List of Revised Required Actions to Corresponding PRA Functions Required Action Condition Description Applicable SSCs SSC Function(s)

PRA Modeled Design Success Criteria PRA Success Criteria Comments Condition E One channel inoperable (FU12)

Two Underfrequency Bus A01, A02 channels per bus Reactor Trip Initiation No 1 of 2 Underfrequency Bus channels on both A01 and A02 Not Modeled - See comments The condition of one of two inoperable reactor trip breakers is used as a surrogate for this condition.

Condition K One channel inoperable (FU7a)

Four Pressurizer Pressure - Low channels Reactor Trip Initiation No 2 of 4 Pressurizer Pressure - Low channels Not Modeled - See comments The condition of one of two inoperable reactor trip breakers is used as a surrogate for this condition.

(FU8)

Three Pressurizer Water Level -

High channels Reactor Trip Initiation No 2 of 3 Pressurizer Water Level - High channels Not Modeled - See comments The condition of one of two inoperable reactor trip breakers is used as a surrogate for this condition.

(FU9b)

Three Reactor Coolant Flow -

Low (Two loops) channels per RCS Loop Reactor Trip Initiation No 2 of 3 Reactor Coolant Flow - Low channels on both RCS Loops when above P8 Not Modeled - See comments The condition of one of two inoperable reactor trip breakers is used as a surrogate for this condition.

(FU11)

Two Undervoltage Bus A01, A02 channels per electrical bus Reactor Trip Initiation No 1 of 2 Undervoltage Bus channels on A01 and A02 Not Modeled - See comments The condition of one of two inoperable reactor trip breakers is used as a surrogate for this condition.

Condition L One Reactor Coolant Flow-Low (Single Loop) channel inoperable (FU9a)

Three Reactor Coolant Flow -

Low (Single Loop) channels per RCS Loop Reactor Trip Initiation No 2 of 3 Reactor Coolant Flow - Low channels on any RCS Loop when below P8 Not Modeled - See comments The condition of one of two inoperable reactor trip breakers is used as a surrogate for this condition.

Condition M One Reactor Coolant Pump Breaker Position (Single Loop) channel inoperable (FU10a)

One RCP Breaker Position (Single Loop) channel per RCP Reactor Trip Initiation No One RCP Breaker Position channel on any RCP when below P8 Not Modeled - See comments The condition of one of two inoperable reactor trip breakers is used as a surrogate for this condition.

Condition N One inoperable channel (FU10b)

One RCP breaker position (two loops) channel per RCP Reactor Trip Initiation No One RCP breaker Position channel on both RCPs when above P8 Not Modeled - See comments The condition of one of two inoperable reactor trip breakers is used as a surrogate for this condition.

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 Page 5 of 27 Table E1 List of Revised Required Actions to Corresponding PRA Functions Required Action Condition Description Applicable SSCs SSC Function(s)

PRA Modeled Design Success Criteria PRA Success Criteria Comments Condition O One turbine trip channel inoperable (FU15a)

Three Turbine trip - Low Auto-Stop Oil Pressure channels Reactor Trip Initiation No 2 of 3 Turbine trip - Low Auto-Stop Oil Pressure channels Not Modeled - See comments The condition of one of two inoperable reactor trip breakers is used as a surrogate for this condition.

(FU15b)

Two Turbine trip - Stop Valve Closure channels Reactor Trip Initiation No 2 of 3 Turbine trip - Stop Valve Closure channels Not Modeled - See comments The condition of one of two inoperable reactor trip breakers is used as a surrogate for this condition.

Condition P One train inoperable (Modes 1,2)

(FU16)

Two SI input from ESFAS trains Reactor Trip Initiation No 1 of 2 SI input from ESFAS trains Not Modeled - See comments The condition of one of two inoperable reactor trip breakers is used as a surrogate for this condition.

(FU21)

Two Automatic Trip Logic trains Reactor Trip Initiation No 1 of 2 Automatic Trip Logic trains Not Modeled - See comments The condition of one of two inoperable reactor trip breakers is used as a surrogate for this condition.

Condition Q One RTB inoperable (Modes 1,2)

(FU18)

Two RTB trains Reactor Trip Initiation No 1 of 2 RTB trains Not Modeled - See comments This SSC is used as a surrogate for other TS 3.3.1 RPS Instrumentation Conditions.

Condition U One trip mechanism inoperable for one RTB (Modes 1,2)

(FU19)

One RTB Undervoltage and Shunt Trip Mechanism per RTB train Reactor Trip Initiation No One RTB Undervoltage and Shunt Trip Mechanism on any RTB train Not Modeled - See comments The condition of one of two inoperable reactor trip breakers is used as a surrogate for this condition.

TS 3.3.2, ESFAS Instrumentation (Table 3.3-2)

Condition B One channel inoperable (FU1a)

Two Manual Initiation channels Safety Injection Initiation No 1 of 2 Manual Initiation channels Not Modeled - See comments The operator actions for failure to manually actuate SI will be used as a surrogate to conservatively bound the risk increase associated with this function as permitted by NEI 06-09.

(FU3a)

Two Manual Initiation channels Containment Isolation Initiation No 1 of 2 Manual Initiation channels Not Modeled - See comments The condition of manual SI function inoperable is used as a surrogate for this condition since an SI signal generates a CI signal.

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 Page 6 of 27 Table E1 List of Revised Required Actions to Corresponding PRA Functions Required Action Condition Description Applicable SSCs SSC Function(s)

PRA Modeled Design Success Criteria PRA Success Criteria Comments Condition C One train inoperable.

(FU1b)

Two Automatic Actuation Logic and Actuation Relay trains Safety Injection Initiation No 1 of 2 Automatic Actuation Logic and Actuation Relay trains channels Not Modeled - See comments The failure of the automatic SI signals will be used as a surrogate to conservatively bound the risk increase associated with this function as permitted by NEI 06-09.

(FU3b)

Two Automatic Actuation Logic and Actuation Relay trains Containment Isolation Initiation No 1 of 2 Automatic Actuation Logic and Actuation Relay trains Not Modeled - See comments The condition of automatic SI function inoperable is used as a surrogate for this condition since an SI signal generates a CI signal.

Condition D One channel inoperable.

(FU1c)

Three Containment Pressure -

High channels Safety Injection Initiation No 2 of 3 Containment Pressure - High channels Not Modeled - See comments The condition of automatic SI function inoperable is used as a surrogate for this condition since it is the same function (SI initiation).

(FU1d)

Three Pressurizer Pressure -

Low channels Safety Injection Initiation No 2 of 3 Pressurizer Pressure - Low channels Not Modeled - See comments The condition of automatic SI function inoperable is used as a surrogate for this condition since it is the same function (SI initiation).

(FU1e)

Three Steam Line Pressure -

Low channels per steam line Safety Injection Initiation No 2 of 3 Steam Line Pressure - Low channels per main steam line Not Modeled - See comments The condition of automatic SI function inoperable is used as a surrogate for this condition since it is the same function (SI initiation).

(FU4c)

Three Containment Pressure -

High, High channels Steam Line Isolation Initiation No 2 of 3 Containment Pressure - High, High channels Not Modeled - See comments The failure of the model logic for steam generator isolation will be used as a surrogate to conservatively bound the risk increase associated with this function as permitted by NEI 06-09.

(FU4d)

Two High Steam Flow channels; Coincident with SI and Coincident with three Tavg - Low, Low channels per RCS loop Steam Line Isolation Initiation No 1 of 2 High Steam Flow channels coincident with SI and coincident with 2 of 3 Tavg - Low, Low channels Not Modeled - See comments The condition of steam generator isolation function inoperable is used as a surrogate for this condition.

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 Page 7 of 27 Table E1 List of Revised Required Actions to Corresponding PRA Functions Required Action Condition Description Applicable SSCs SSC Function(s)

PRA Modeled Design Success Criteria PRA Success Criteria Comments (FU4e)

Two High, High Steam Flow channels per steam line; Coincident with SI Steam Line Isolation Initiation No 1 of 2 High, High Steam Flow channels per steam line coincident with SI Not Modeled - See comments The condition of steam generator isolation function inoperable is used as a surrogate for this condition.

(FU5b)

Three SG Water Level - High channels per SG Feedwater Isolation Initiation No 2 of 3 SG Water Level -

High channels per SG Not Modeled - See comments The condition of steam generator isolation function inoperable is used as a surrogate for this condition.

(FU6b)

Three SG Water Level - Low, Low channels per SG Auxiliary Feedwater Initiation No 2 of 3 SG Water Level -

Low, Low channels on any SG Not Modeled - See comments The failure of the model logic for these relays will be used as a surrogate to conservatively bound the risk increase associated with this function as permitted by NEI 06-09.

Condition F One channel inoperable (FU4a)

One Manual Initiation channel per RCS loop Steam Line Isolation Initiation No One Manual Initiation channel per RCS loop Not Modeled - See comments The condition of steam generator isolation function inoperable is used as a surrogate for this condition.

Condition G One train inoperable (FU4b)

Two Automatic Actuation Logic and Actuation Relay trains Steam Line Isolation Initiation No 1 of 2 Automatic Actuation Logic and Actuation Relay trains Not Modeled - See comments The condition of steam generator isolation function inoperable is used as a surrogate for this condition.

(FU5a)

Two Automatic Actuation Logic and Actuation Relay trains Feedwater Isolation No 1 of 2 Automatic Actuation Logic and Actuation Relay trains Not Modeled - See comments The condition of AFW initiation is used as a surrogate for this condition.

(FU6a)

Two Automatic Actuation Logic and Actuation Relay trains Auxiliary Feedwater No 1 of 2 Automatic Actuation Logic and Actuation Relay trains Not Modeled - See comments The failure of the model logic for these relays will be used as a surrogate to conservatively bound the risk increase associated with this function as permitted by NEI 06-09.

Condition H One channel inoperable (FU6d)

Two Undervoltage Bus A01 and A02 channels per electrical bus Auxiliary Feedwater No 1 of 2 Undervoltage Bus channels on A01 and A02 Not Modeled - See comments The failure of the model logic for starting all four AFW pumps will be used as a surrogate to conservatively bound the risk increase associated with this function as permitted by NEI 06-09.

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 Page 8 of 27 Table E1 List of Revised Required Actions to Corresponding PRA Functions Required Action Condition Description Applicable SSCs SSC Function(s)

PRA Modeled Design Success Criteria PRA Success Criteria Comments TS 3.4, Reactor Coolant System TS 3.4.11 Condition B One PORV inoperable and not capable of being manually cycled Two Power Operated Relief Valves (PORVs)

RCS pressure control Yes 1 of 2 PORVs Same TS 3.4.11 Condition C One block valve inoperable Two PORV Block Valves RCS integrity Yes Associated Block Valve closure Same TS 3.5, Emergency Core Cooling System (ECCS)

TS 3.5.2 Condition A One ECCS train inoperable Two ECCS trains each comprised of one SI pump, one RHR pump, one RHR heat exchanger and associated RWST and containment sump flowpaths Emergency core cooling and post-accident (long-term) core cooling Yes 1 of 2 SI pumps, and 1 of 2 RHR pumps 1 of 2 RHR pumps w/ suction from containment sump, supplying suction to 1 of 2 SI pumps for flowpath to RCS Same TS 3.6, Containment Systems TS 3.6.2 Condition C One or more containment air locks inoperable for reasons other Condition A or B One equipment hatch; One personnel airlock; Two emergency airlocks Containment integrity No One of two containment air lock doors closed with acceptable containment leakage Not Modeled - See comments The failure of the model logic for containment penetrations will be used as a surrogate to conservatively bound the risk increase associated with this function as permitted by NEI 06-09.

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 Page 9 of 27 Table E1 List of Revised Required Actions to Corresponding PRA Functions Required Action Condition Description Applicable SSCs SSC Function(s)

PRA Modeled Design Success Criteria PRA Success Criteria Comments TS 3.6.3 Condition A One or more penetration flow paths with one containment isolation valve inoperable (applicable to penetration flow paths with two containment isolation valves)

Two isolation valves on each containment penetration Containment integrity No 1 of 2 isolation valves per penetration isolate Not Modeled - See comments The failure of the model logic for containment penetrations will be used as a surrogate to conservatively bound the risk increase associated with this function as permitted by NEI 06-09.

TS 3.6.3 Condition C One or more penetration flow paths with one containment isolation valve inoperable (applicable to penetration flow paths with one containment isolation valve and a closed system)

One isolation valve and one closed system on each containment penetration Containment integrity No Each isolation valve per penetration isolates Not Modeled - See comments The failure of the model logic for containment penetrations will be used as a surrogate to conservatively bound the risk increase associated with this function as permitted by NEI 06-09.

TS 3.7, Plant Systems TS 3.7.2 Condition A One Steam Generator flowpath with one or more inoperable valves in MODE 1 Two Main Steam Lines equipped with one Main Steam Isolation Valve (MSIV) and one Non-Return Check Valve Steam Line Isolation from Faulted Steam Line Yes MSIV on affected steam line isolates Same

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 Page 10 of 27 Table E1 List of Revised Required Actions to Corresponding PRA Functions Required Action Condition Description Applicable SSCs SSC Function(s)

PRA Modeled Design Success Criteria PRA Success Criteria Comments TS 3.7.4 Condition A One required ADV flowpath inoperable Two Atmospheric Dump Valves (ADVs)

Facilitate Unit cooldown to SDC conditions Yes 1 of 2 ADV flowpaths Same TS 3.7.5 Condition A Turbine driven AFW pump system inoperable due to one inoperable steam supply, OR Turbine driven AFW pump system inoperable in MODE 3 following refueling One turbine-driven AFW pump and one motor-driven AFW pump and associated CST and SW suction piping Feedwater supply to SGs upon loss of main feedwater Yes One motor driven AFW pump supplies CST feedwater to both SGs Same TS 3.7.5 Condition B One AFW pump system inoperable in MODE 1, 2 or 3 for reasons other than Condition A One turbine-driven AFW pump and one motor-driven AFW pump and associated CST and SW suction piping Feedwater supply to SGs upon loss of main feedwater Yes One turbine driven AFW pump supplies CST feedwater to both SGs Same TS 3.7.7 Condition A One CC pump inoperable Two Component Cooling Water (CCW) trains each consisting of one CCW pump and one CCW heat exchanger, and one common CCW heat exchanger capable of aligning to either train Heat sink for removing process and operating heat from safety-related components Yes 1 CCW pump and 1 CCW HX provide heat sink to safety related equipment Same

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 Page 11 of 27 Table E1 List of Revised Required Actions to Corresponding PRA Functions Required Action Condition Description Applicable SSCs SSC Function(s)

PRA Modeled Design Success Criteria PRA Success Criteria Comments TS 3.7.7 Condition B One required CC heat exchanger inoperable Two Component Cooling Water (CCW) trains each consisting of one CCW pump and one CCW heat exchanger, and one common CCW heat exchanger capable of aligning to either train Heat sink for removing process and operating heat from safety-related components Yes 1 CCW pump and 1 CCW HX provide heat sink to safety related equipment Same TS 3.7.8 Condition A One SW pump inoperable AND Both units in Modes 1, 2, 3, or 4 Six Service Water (SW) system pumps, one common ring header, non-essential flowpath isolation valves and associated SW intake piping Heat sink for removing process and operating heat from safety-related components Yes 2 SW pumps and 1 SW Ring Header provide heat sink to CCW system and essential loads; auto-isolate non-essential flowpaths Same TS 3.7.8 Condition C SW ring header continuous flowpath interrupted Six Service Water (SW) system pumps, one common ring header, non-essential flowpath isolation valves and associated SW intake piping Heat sink for removing process and operating heat from safety-related components Yes 2 SW pumps and 1 SW Ring Header provide heat sink to CCW system and essential loads; auto-isolates non-essential flowpaths Same TS 3.7.8 Condition D One or more non-essential-SW-load flowpath(s) with one required automatic isolation valve inoperable.

AND Affected non-essential flowpath(s) not isolated Six Service Water (SW) system pumps, one common ring header, non-essential flowpath isolation valves and associated SW intake piping Heat sink for removing process and operating heat from safety-related components Yes 3 SW pumps and 1 SW Ring Header provide heat sink to CCW system and essential loads; auto-isolates non-essential flowpaths Same

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 Page 12 of 27 Table E1 List of Revised Required Actions to Corresponding PRA Functions Required Action Condition Description Applicable SSCs SSC Function(s)

PRA Modeled Design Success Criteria PRA Success Criteria Comments TS 3.8, Electrical Power Systems TS 3.8.1 Condition A Associated unit 345/13.8 kV (X03) transformer inoperable.

OR Gas turbine not in operation when utilizing opposite units 345/13.8 kV (X03) transformer.

Two 13.8 kV station auxiliary transformers; one 13.8 kV gas turbine; two Class 1E 4.16 kV electrical buses; four 4.16 kV diesel generators (DG)

Power onsite safeguards buses from offsite and onsite transmission networks to support normal, safe shutdown and accident mitigation conditions Yes Automatically power associated ESF busses Same TS 3.8.1 Condition B Associated unit's 13.8/4.16kV (X04) transformer inoperable Two 13.8 kV station auxiliary transformers; one 13.8 kV gas turbine; two Class 1E 4.16 kV electrical buses; four 4.16 kV diesel generators (DG)

Power onsite safeguards buses from offsite and onsite transmission networks to support normal, safe shutdown and accident mitigation conditions Yes Automatically power associated ESF busses Same

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 Page 13 of 27 Table E1 List of Revised Required Actions to Corresponding PRA Functions Required Action Condition Description Applicable SSCs SSC Function(s)

PRA Modeled Design Success Criteria PRA Success Criteria Comments TS 3.8.1 Condition C Associated unit's required offsite power source to buses A05 and A06 inoperable.

OR Required offsite power source to buses 1A05 and 2A06 inoperable Two 13.8 kV station auxiliary transformers; one 13.8 kV gas turbine; two Class 1E 4.16 kV electrical buses; four 4.16 kV diesel generators (DG)

Power onsite safeguards buses from offsite and onsite transmission networks to support normal, safe shutdown and accident mitigation conditions Yes Automatically power associated ESF busses Same TS 3.8.1 Condition D One or more required offsite power source(s) to one or more required Class 1E 4.16 kV bus(es) inoperable Two 13.8 kV station auxiliary transformers; one 13.8 kV gas turbine; two Class 1E 4.16 kV electrical buses; four 4.16 kV diesel generators (DG)

Power onsite safeguards buses from offsite and onsite transmission networks to support normal, safe shutdown and accident mitigation conditions Yes Automatically power associated ESF busses Same TS 3.8.1 Condition F One or more required offsite power source to one or more Class 1E 4.16 kV safeguards bus(es) inoperable AND Standby emergency power inoperable to Two 13.8 kV station auxiliary transformers; one 13.8 kV gas turbine; two Class 1E 4.16 kV electrical buses; four 4.16 kV diesel generators (DG)

Power onsite safeguards buses from offsite and onsite transmission networks to support normal, safe shutdown and accident Yes Automatically power associated ESF busses Same

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 Page 14 of 27 Table E1 List of Revised Required Actions to Corresponding PRA Functions Required Action Condition Description Applicable SSCs SSC Function(s)

PRA Modeled Design Success Criteria PRA Success Criteria Comments redundant equipment mitigation conditions TS 3.8.4 Condition A One DC electrical power subsystem inoperable 4 battery banks and associated chargers and motor control centers (MCC)

Ensure availability of required DC power to shut down the reactor and maintain it in a safe condition Yes Automatically power associated ESF busses Same TS 3.8.7 Condition A One required inverter inoperable Four 120 VAC Instrument Inverters Ensure availability of required DC power to shut down the reactor and maintain it in a safe condition Yes Two 120 VAC Instrument Inverters per electrical train Same

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Page 15 of 27 2.0 MODEL CONFIRMATION Table E1-2 provides the calculated RICT for each individual condition to which the RICT applies (assuming no other structures, systems, and components (SSCs) modeled in the PRA are unavailable). Table E1-2 confirms that the PRA models include the necessary SSCs and their functions to address each proposed application of the RICT Program to the TS Required Actions. The RICT estimates are based on the internal events, internal flooding, and internal fire PRA model calculations with seismic CDF and LERF penalties and are the most limiting of the Point Beach Unit 1 and Unit 2 results. Actual RICT values will be calculated based on the actual plant configuration using a current revision of the PRA model which represents the as-built, as-operated condition of the plant, as required by NEI 06-09-A (Reference 1) and the NRC safety evaluation (Reference 2), and may differ from the RICTs presented in Table E1-2 below. RICTs calculated to be greater than 30 days are capped at 30 days based on NEI 06-09-A. RICTs not capped at 30 days are rounded to nearest number of days. TS Required Action Conditions with insufficient TS operable equipment to meet the specified safety function of the system are not eligible for RICT Program application.

Consistent with NEI 06-09-A, cases where the total CDF or LERF is greater than 1E-03/year or 1E-04/year, respectively, are not eligible for RICT Program application.

References:

1. NRC letter to NEI, Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, Risk-Informed Technical Specifications Initiative 4B, Risk-Managed Technical Specifications (RMTS) Guidelines (TAC No. MD4995), dated May 17, 2007 (ADAMS Accession No. ML071200238)

2. NRC letter to NEI, Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, Risk-Informed Technical Specifications Initiative 4B, Risk-Managed Technical Specifications (RMTS) Guidelines (TAC No. MD4995), dated May 17, 2007 (ADAMS Accession No. ML071200238)

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Page 16 of 27 Table E1 In-Scope TS/LCO Conditions RICT Estimate TS/LCO Required Action Condition Description RICT Estimate (days)

Reactor Protection System (RPS) Instrumentation 3.3.1.B One Manual Reactor Trip channel inoperable 30 3.3.1.D One channel inoperable 30 3.3.1.E One channel inoperable 30 3.3.1.K One channel inoperable 30 3.3.1.L One Reactor Coolant Flow-Low (Single Loop) channel inoperable 30 3.3.1.M One Reactor Coolant Pump Breaker Position, (Single Loop) channel inoperable 30 3.3.1.N One channel inoperable 30 3.3.1.O One turbine trip channel, inoperable 30 3.3.1.P One train inoperable, (Modes 1,2) 30 3.3.1.Q One RTB inoperable, (Modes 1,2) 30 3.3.1.U One trip mechanism, inoperable for one RTB (Modes 1,2) 30 Engineered Safety Feature Actuation System (ESFAS) Instrumentation 3.3.2.B One channel inoperable 30 3.3.2.C One channel inoperable 30 3.3.2.D One channel inoperable 30 3.3.2.F One channel inoperable 30 3.3.2.G One train inoperable 30 3.3.2.H One channel inoperable 18 Pressurizer Power Operated Relief Valves (PORVs) 3.4.11.B One PORV inoperable and not capable of being manually cycled 30 3.4.11.C One block valve, inoperable 30 ECCS - Operating 3.5.2.A One ECCS train inoperable 30 Containment Air Locks 3.6.2.C One or more, containment air locks inoperable for reasons other than Condition A or B 9

Containment Isolation Valves 3.6.3.A One or more penetration flow paths with one containment isolation valve inoperable 9

3.6.3.C One or more penetration flow paths with one containment isolation valve inoperable 9

Main Steam Isolation Valves (MSIVs) and Non-Return Check Valves 3.7.2.A One Steam Generator, flowpath with one or more inoperable valves in MODE 1 30 Atmospheric Dump Valve (ADV) Flowpaths 3.7.4.A One required ADV flowpath inoperable 30 Auxiliary Feedwater (AFW) 3.7.5.A Turbine driven AFW pump system inoperable due to one inoperable steam supply, OR Turbine driven AFW pump system inoperable in MODE 3 following refueling 30 3.7.5.B One AFW pump system inoperable in MODE 1, 2 or 3 for reasons other than Condition A 30

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Page 17 of 27 Table E1 In-Scope TS/LCO Conditions RICT Estimate TS/LCO Required Action Condition Description RICT Estimate (days)

Component Cooling Water (CC) System 3.7.7.A One CC pump, inoperable 23 3.7.7.B One required CC heat exchanger inoperable 30 Service Water (SW) System 3.7.8.A One SW pump inoperable AND Both units in Modes 1, 2, 3, or 4 30 3.7.8.C SW ring header, continuous flowpath, interrupted 30 3.7.8.D One or more non-essential-SW-load flowpath(s) with one required automatic isolation valve inoperable AND Affected non-essential flowpath(s),

not isolated.

30 AC Sources - Operating 3.8.1.A Associated unit 345/13.8 kV (X03) transformer inoperable OR Gas turbine not in operation when utilizing opposite units 345/13.8 kV (X03) transformer.

30 3.8.1.B Associated unit's 13.8/4.16kV (X04) transformer, inoperable 30 3.8.1.C Associated unit's required offsite power source to buses A05 and A06 inoperable OR Required offsite power source to buses 1A05 and 2A06 inoperable.

30 3.8.1.D One or more required offsite power source(s) to one or more required Class 1E 4.16 kV bus(es) inoperable.

30 3.8.1.F One or more required offsite power source to one or more Class 1E 4.16 kV safeguards bus(es) inoperable AND Standby emergency power inoperable to redundant equipment.

8 DC Sources - Operating 3.8.4.A One DC electrical power subsystem inoperable 7

Inverters - Operating 3.8.7.A One required inverter inoperable 30

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Page 18 of 27 3.0 ADDITIONAL JUSTIFICATION FOR SPECIFIC ACTIONS The NRC Final Revised Model Safety Evaluation for TSTF-505, Revision 2 (Reference 1) provides a listing of TS Conditions for inclusion in licensee RICT Programs that are recommended for additional technical justification, as specified in Table 1, Conditions Requiring Additional Technical Justification, for NUREG-1431, Westinghouse STS plants. The discussion below addresses the Point Beach TS Conditions that are applicable to the NRCs additional technical justification recommendations along with the bases for the conclusion that the proposed changes do not result in any loss of a specified safety function.

1. TS 3.3.1, Table 3.3.1-1, FU 2a, Power Range Neutron Flux - High channels Condition D: One channel inoperable.

Required ACTION: Place channel in trip Four power range neutron flux - high channels are provided for overpower protection. The trip logic accounts for the power range nuclear flux - high channels also providing a rod control function whereby the system must be capable of withstanding a second channel failure to satisfy the single failure criterion. The 2 out of 4 trip logic defaults to a 2 out of 3 logic in the event of a failed or bypassed power range neutron flux-high channel. The unavailability of a single power range neutron flux-high channel will not impair reactor core power distribution monitoring or control rod regulation since the rod control signals are based on the average of the power range neutron flux-high signals and the rod control system only responds to rapid changes in neutron flux. All neutron flux power range currents are indicated in the control room. If a power range channel failure occurs, switches are provided to permit the failed power range channel's overpower rod stop function to be bypassed and its average power signal to the reactor control system replaced by a signal derived from an active channel. This allows normal power operation to continue while the failed channel is repaired. Alarms are also provided to alert the operator of deviations from normal operating conditions so that corrective action can be taken prior to reaching a reactor trip setting. Thereby, the inoperability of a power range neutron flux-high channel does not result in a loss of function.

2. TS 3.3.1, Table 3.3.1-1, FU 18, Condition Q for the Reactor Trip Breakers (RTBs), MODES 1, 2 Condition Q: One RTB inoperable.

Required ACTION: Restore RTB to OPERABLE status An RTB train consists of all trip breakers associated with a single RTS logic train that are racked in and capable of supplying power to the rod control system. Two devices in each breaker receive RPS signals, either of which will trip the RTBs via an undervoltage trip device which will trip the reactor on loss of breaker control power or the receipt of an RPS trip signal, or a shunt trip device which provides a backup if the passive undervoltage trip fails to open the reactor trip breakers upon receipt of a trip signal. In addition, an Anticipated Transient Without Scram Mitigating System Actuation Circuitry (AMSAC) system is provided in the event of an anticipated trip without scram (ATWS) condition to further mitigate the effects of a failed RPS by tripping the turbine and starting AFW in addition to inserting the control rods. Thereby, sufficient redundancy and diversity exists in the RTB system to assure the unavailability of a single RTB does not result in a loss of function.

3. TS 3.5.2, ECCS - Operating Condition A: One ECCS train inoperable.

Required ACTION: Restore train to OPERABLE status Additional justification is not needed for this TS Condition since Point Beach TS 3.5.2, Condition A, addresses only one inoperable ECCS train.

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Page 19 of 27

4. TS 3.6.2, Containment Air Locks Condition C: One or more containment air locks inoperable for reasons other than Condition A or B.

Required ACTION:

C.1 Initiate action to evaluate overall containment leakage rate per LCO 3.6.1, AND C.2 Verify a bulkhead door and associated equalizing valve are closed in the affected air lock, AND C.3 Restore air lock to OPERABLE status The containment airlocks are comprised of a containment penetration chamber isolated by inner and outer bulkheads equipped with double o-ring seals. In the event of an inoperable containment airlock, ACTION C.1 requires the condition to be immediately assessed in accordance with TS 3.6.1 (i.e.,

immediately initiate action to evaluate overall primary containment leakage). In addition, compliance TS 3.6.2, ACTION C.2, assures the presence of at least one physical barrier on the affected airlock penetration with acceptable barrier leakage in accordance with the Point Beach Containment Leakage Rate Testing Program. Thereby, containment integrity function is maintained throughout the period of airlock inoperability consistent with plant safety analyses.

5. TS 3.7.2, Main Steam Isolation Valves (MSIVs) and Non-Return Check Valves Condition A: One Steam Generator flowpath with one or more inoperable valves in MODE 1 Required ACTION: Restore valve to OPERABLE status.

TS 3.7.2 specifies requirements for the MSIVs, which mitigate a steam line break by automatically closing upon receipt of a safety injection (SI) signal coincident with a Hi-Hi steam flow signal. Rapid MSIV closure limits RCS cooldown and the resulting reactivity addition. Automatic MSIV closure also prevents the uncontrolled blowdown of a SG resulting from a downstream steam line break. In addition, TS 3.7.2 specifies requirements for the main steam non-return check valves, which close on reverse flow to prevent the blowdown of a non-faulted SG in the event of a main steam line break coincident with the failure of a MSIV to close properly. Main steam non-return check valve closure assures that for any steam break location coupled with an MSIV single failure, both SGs will not blow down and cause core damage due to excess positive reactivity from the RCS cooldown transient. Thereby, the loss of a single MSIV will not result in a loss of the steam line isolation function.

References:

1.

NRC Safety Evaluation, Final Revised Model Safety Evaluation of Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, dated November 21, 2018 (ADAMS Accession No. ML18269A041) 4.0 INFORMATION TO SUPPORT INSTRUMENTATION REDUNDANCY AND DIVERSITY The NRCs revised model application for TSTF 505, Revision 2 (Reference 1), recommends for the proposed changes to the protective instrumentation features in TS Section 3.3, "Instrumentation," a demonstration that at least one redundant or diverse means remains available to accomplish the associated safety function(s) during application of a RICT, consistent with the defense-in-depth philosophy of Regulatory Guide (RG) 1.174 (Reference 2). The request is in recognition that while in an ACTION statement, redundancy of the protective feature, and thereby system reliability, is temporarily reduced.

Table E1.1 of this enclosure provides a description of the PRA modeling for RPS and ESFAS instrumentation, including the scope of the TS functions to which a RICT would be applied. Sections 4.1 and 4.2 below demonstrate that diversity and redundancy is maintained during the application of a RICT by demonstrating the existence of at least one additional means to accomplish the safety function(s).

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Page 20 of 27 4.1 TS 3.3.1 - Reactor Protection System (RPS) Instrumentation The Point Beach RPS consists of four instrument channels that monitor up to four various plant parameters, depending on the coincidence logic required for the specific trip. Each protection channel terminates at a channel trip bistable in the analog protection racks. Each channel trip bistable controls two independent and redundant logic relays associated with the two independent and redundant trains (A and B). The logic relays for each train are combined in a coincidence logic network (e.g., two-out-of-four). Two independent and redundant reactor trip breakers in series provide power to the control rod drive mechanisms. In addition, two independent and redundant bypass breakers are provided in parallel with the reactor trip breakers to allow for continued reactor operation during testing of the reactor trip breakers.

When the required number of channels (e.g., two-out-of-four) indicate that a plant parameter is outside its acceptable operating limit, their associated channel bistables are tripped. The tripping of the channel bistables result in the tripping of their associated coincidence logic relays for each train, which in turn results in de-energizing the reactor trip relays. De-energizing the reactor trip relays causes the associated train trip breaker to open by de-energizing its undervoltage trip coil and by energizing its shunt trip coil through an interposing relay. De-energizing the reactor trip relays also causes the opposite train bypass breaker to open by de-energizing its undervoltage trip coil. When the reactor trip breakers are tripped, power to the control rod drive mechanisms is interrupted, which allows the control rods to insert into the core by gravity.

The RPS is designed so that no single failure within, or in an associated system which supports RPS operation, will prevent the intended reactor trip function. The RPS is redundant and independent for all primary inputs and functions. Each RPS channel is functionally independent of every other RPS channel and receives power from a separate AC power source. Train separation is achieved by providing separate racks and each train receives power from a separate DC power source such that each RPS train is functionally independent of the redundant train. The RPS is designed to IEEE Standard 279-1968 (Reference 3) except for some backup/anticipatory reactor protection which may not fully conform to all IEEE 279 criteria, as described in the Point Beach UFSAR.

The primary reactor trip functions are the overpower T, overtemperature T, and nuclear overpower trips, which define the allowable region of reactor power and coolant temperature conditions. The high pressurizer water level, loss of RCS flow, steam and feedwater flow mismatch, steam generator low-low level, turbine, safety injection, nuclear source and intermediate range, and manual trip functions are provided to back up the primary tripping functions for specific accident conditions and mechanical failures.

RPS diversity and redundancy in trip actuation capability is depicted in Table E1-3(1) below.

References:

1. NRC Revised TSTF 505, Revision 2, Model Application (ADAMS Accession No. ML18115A482)

2. Regulatory Guide 1.174. An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, January 2018 (ADAMS Accession No. ML17317A256)

3. IEEE 279, Proposed IEEE Criteria for Nuclear Power Plant Protection Systems, dated August 1968.

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 Page 21 of 27 Table E1-3(1) - Information to Support RPS Instrumentation Redundancy and Diversity RPS Function Functional Unit (FU)

Channels to Trip (1)(2)(3)

Primary Transient /

Accident Protection (4)

Diverse Actuation Functions() (1)

Comments Power Range Neutron Flux (high setting)

FU2a 2 out of 4

- Uncontrolled RCCA Withdrawal at Power

- Uncontrolled RCCA Withdrawal from a Subcritical Condition

- Steam Line Break

- Loss of External Electrical Load

- Loss of RCS Coolant

- Overtemperature T

- Overpower T

- Pressurizer Pressure - High

- Pressurizer Level - High

- Manual Trip No interlocks (i.e. always active)

Power Range Neutron Flux (low setting)

FU2b 2 out of 4

- Uncontrolled RCCA Withdrawal from a Subcritical Condition

- Loss of RCS Coolant

- Source Range Flux

- Intermediate Range Flux

- Power Range Neutron Flux - High

- Manual Trip Manual block above P10 interlock (~10% RTP) allowed. Automatically reinstated below P10.

Overtemperature T FU5 2 out of 4

- Uncontrolled RCCA Withdrawal from a Subcritical Condition

- Uncontrolled RCCA Withdrawal at Power RCCA Drop

- Loss of RCS Flow

- RCCA Drop

- Loss of External Electrical Load

- Loss of Normal Feedwater

- SG Tube Rupture

- Loss of RCS Coolant

- Power Range Neutron Flux - High

- Overpower T

- Pressurizer Pressure - High

- Pressurizer Level - High

- SG Level Low-Low

- Turbine Trip

- Manual Trip No interlocks. Requires two sets of temperature measurements above trip setpoint per RCS loop Overpower T FU6 2 out of 4

- Uncontrolled RCCA Withdrawal at Power

- RCCA Drop

- Steam Line Break

- CVCS Malfunction

- Source Range Neutron Flux

- Manual Trip No interlocks. Requires two sets of temperature measurements above trip setpoint per RCS loop

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 Page 22 of 27 Table E1-3(1) - Information to Support RPS Instrumentation Redundancy and Diversity RPS Function Functional Unit (FU)

Channels to Trip (1)(2)(3)

Primary Transient /

Accident Protection (4)

Diverse Actuation Functions() (1)

Comments Pressurizer Pressure -

High FU7a 2 out of 3

- Uncontrolled RCCA Withdrawal at Power

- Loss of External Electrical Load

- Loss of RCS Flow

- Loss of Normal Feedwater

- Power Range Neutron Flux - High

- Overtemperature T

- Pressurizer Level - High

- SG Level Low-Low

- Turbine Trip

- Manual Trip No interlocks Pressurizer Pressure -

Low FU7b 2 out of 4

- RCCA Drop

- SG Tube Rupture

- Steam Line Break

- Loss of RCS Coolant

- Overtemperature T

- Overpower T

- Safety Injection

- Manual Trip Automatic block below P-7 Reinstated above P-7 Pressurizer Level - High FU8 2 out of 3

- Uncontrolled RCCA Withdrawal at Power

- Loss of External Electrical Load

- Loss of Normal Feedwater

- Pressurizer Pressure - High

- Power Range Neutron Flux - High

- Overtemperature T

- Manual Trip Automatic block below P-7 RCS Flow - Low (one loop) (two loops)

FU9a FU9b 2 out of 3

- Loss of RCS Flow

- Loss of All AC Power to the Auxiliaries

- Overtemperature T

- Pressurizer Pressure - High

- Undervoltage Bus A01, A02

- Underfrequency Bus A01, A02

- RCP Breaker position

- Manual Trip One loop trip requires 2 of 3 channels on either RCS loop.

Blocked below P8 interlock

(~35% RTP). Reinstated above P8.

Two loop trip requires 2 of 3 channels on both RCS loops.

Blocked below P7 (~10%

RTP). Reinstated above P7.

RCP breaker position -

one loop RCP breaker position -

2 loops FU10a FU10b 1 out of 1 per loop

- Loss of RCS Flow

- Loss of All AC Power to the Auxiliaries

- Overtemperature T

- Pressurizer Pressure - High

- Undervoltage Bus A01, A02

- RCS Flow - Low (one loop) (two loops)

- Manual Trip No interlocks. ~5 second time delay. Trips RCP on affected bus. One loop interlocked with P-8. Two loops interlocked with P-7.

Undervoltage Bus A01, A02 FU11 1 out of 2 per bus

- Loss of RCS Flow

- Loss of All AC Power to the Auxiliaries

- Overtemperature T

- Pressurizer Pressure - High

- RCP Breaker position

- RCS Flow - Low (one loop) (two loops)

- Manual Trip Automatic block below P-7

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 Page 23 of 27 Table E1-3(1) - Information to Support RPS Instrumentation Redundancy and Diversity RPS Function Functional Unit (FU)

Channels to Trip (1)(2)(3)

Primary Transient /

Accident Protection (4)

Diverse Actuation Functions() (1)

Comments Underfrequency Bus A01, A02 FU12 1 out of 1 per bus

- Loss of RCS Flow

- Loss of All AC Power to the Auxiliaries

- Open RCP Breaker position

- Manual Trip No interlocks. Trips RCP on affected bus.

SG Water Level - Low, Low FU13 2 out of 3

- Loss of Normal Feedwater

- Loss of External Electrical Load

- Loss of All AC Power to the Auxiliaries

- Steam/Feed flow mismatch coincident w/ low SG level

- Overtemperature T

- Overpressure T

- Pressurizer Pressure - High

- Pressurizer Level - High

- RCS Flow - Low

- Undervoltage Bus A01, A02

- Underfrequency Bus A01, A02

- Turbine Trip

- Manual Trip Requires 2 of 3 low-low level channels in either SG SG Water Level Low; coincident w/ Steam Flow

/ Feed Flow Mismatch FU14 1 out of 2 per loop*

- Loss of Normal Feedwater

- Loss of All AC Power to the Auxiliaries

- SG Water Level - Low, Low

- Overtemperature T

- Overpressure T

- Pressurizer Pressure - High

- Pressurizer Level - High

- RCS Flow - Low

- Undervoltage Bus A01, A02

- Underfrequency Bus A01, A02

- Turbine Trip

- Manual Trip

Coincident w/1 out of 2 low SG water level per loop Turbine trip on low oil pressure FU15a 2 out of 3

- Loss of Normal Feedwater

- Loss of External Electrical Load

- SG Water Level - Low, Low

- Pressurizer Pressure - High

- Overtemperature T

- Overpower T

- Undervoltage Bus A01, A02

- Underfrequency Bus A01, A02

- Manual Trip Interlocked with P-7 and P-9

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 Page 24 of 27 Table E1-3(1) - Information to Support RPS Instrumentation Redundancy and Diversity RPS Function Functional Unit (FU)

Channels to Trip (1)(2)(3)

Primary Transient /

Accident Protection (4)

Diverse Actuation Functions() (1)

Comments Notes:

1.

Each Functional Unit (FU) will cause a reactor trip with 1/2, 2/3 or 2/4 tripped signals.

2.

Bypassed channels reduce the number of total available channels by 1, e.g. from 2/4 to 2/3, or from 2/3 to 2/2.

3.

An inoperable channel may be placed in a tripped state, reducing the redundancy from 2/4 required tripped channels to 1/3 required tripped channels.

4.

Each listed accident results in a reactor trip.

Turbine trip on stop valve closure FU15b 2 out of 2

- Loss of Normal Feedwater

- Loss of External Electrical Load

- SG Water Level - Low, Low

- Pressurizer Pressure - High

- Overtemperature T

- Overpower T

- Undervoltage Bus A01, A02

- Underfrequency Bus A01, A02

- Manual Trip Interlocked with P-7 and P-9

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Page 25 of 27 4.2 TS 3.3.2 - Engineering Safeguards Features Actuation System (ESFAS) Instrumentation The Point Beach ESFAS is comprised of sensors connected to signal processing circuitry consisting of two to four redundant channels that monitor various plant parameters and digital circuitry consisting of two redundant logic trains which receive inputs from the signal processing channels. The two ESFAS actuation trains are provided to actuate the two ESF equipment trains associated with each unit. When coincidence logic for a particular ESFAS subsystem is satisfied, the trip relays in both logic trains are actuated. With redundant logic trains, a single logic train failure will not prevent a valid ESF actuation.

The analog portion of the ESFAS system is shared with the RPS. The four ESFAS channels, which share cabinets with the RPS channels, receive 120 VAC power from the four independent, battery-backed instrument buses. The logic racks for the two ESFAS trains that actuate ESF equipment receive battery-backed power from redundant 125 VDC sources. Loss of AC power to an individual ESFAS channel (except the containment spray actuation channels) will cause the associated channel's output bistables to trip. The deenergize-to-operate design is similar to the RPS analog channels. The analog channels are designed to fail in the trip state on a power failure with the exception of the containment spray actuation channels to preclude inadvertent spray-down of the containment on power loss. Similarly, the ESFAS output relays to individual ESF components are intentionally designed as energize-to-trip relays to avoid inadvertent actuation of ESF systems, which would disrupt plant operation.

Four steam line flow channels are used to initiate the Steam Line Isolation (SLI) logic. The four channels are divided into two pairs, with one pair assigned to each main steam line (generator). The flow signals from each main steam line are combined in a 1-out-of-2 coincidence.

Two channels of 4160V bus undervoltage relays are used to initiate the AFW pumps. Each channel contains two relays, which are combined in 1-out-of-2 coincidence. The 1-out-of-2 coincidence outputs are then combined in a taken-twice logic The coincidence logic satisfies the single failure criterion, whereby one relay will not cause unnecessary actuation, and the taken-twice logic requires a loss of both 4160V buses before AFW actuation will occur.

Three channels are used for all other ESFAS trip variables. Three sensor channels support 2-out-of-3 coincidence trip logic, which satisfies both single failure and the reliability criterion. The 2-out-of-3 logic also allows a channel in test to be tripped, while the remaining two channels provide protection in a 1-out-of-2 logic until the tested channel is restored to service The Point Beach ESFAS is designed to IEEE Standard 279-1968 (Reference 1) with the exception of some engineered safety functions which contain equipment that are not credited in accident analyses and may not fully conform to all IEEE 279 criteria, as described in the Point Beach UFSAR.

ESFAS diversity and redundancy in trip actuation capability is depicted in Table E1-3(2) below:

Reference:

1. NRC IEEE 279, Proposed IEEE Criteria for Nuclear Power Plant Protection Systems, dated August 1968.

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 Page 26 of 27 Table E1-3(2) - Information to Support ESFAS Instrumentation Redundancy and Diversity ESFAS Function Functional Unit (FU)

Channels to Trip ESF Actuation Primary Design Basis Accident Diverse Actuation Function(s)

Comments Containment Pressure -

High FU1c 2-out-of-3 (1)

Safety Injection

- Large Break LOCA

- Small Break LOCA

- Pressurizer Pressure - Low SI Pressurizer Pressure -

Low FU1d 2-out-of-3 Safety Injection

- Large Break LOCA

- Small Break LOCA

- SG Tube Rupture

- Containment Pressure - High SI

- SG Water Level - High, High SG Pressure - Low, Low FU1e 2-out-of-3 Safety Injection

- Steam Line Break

- Pressurizer Pressure - Low SI 2-out-of-3 in either loop Containment Pressure -

High, High FU4c 2-out-of-3 Steam Line Isolation

- Small Break LOCA

- Large Break LOCA

- Steam Line Break

- Pressurizer Pressure - Low SI

- Containment Pressure - High SI

- Overpower T SG Flow - High, coincident with SI and Low Tavg FU4d 1-out-of-2 taken twice Steam Line Isolation

- Steam Line Break

- Pressurizer Pressure - Low SI

- Containment Pressure - High SI

- Overpower T 1-out-of-2 in either loop and 2-out-of-4 low Tavg channels. Four steam flow channels are divided into two pairs, with one pair assigned to each steam loop. Flow signals from each loop are combined in a 1-out-of-2 coincidence logic.

SG Flow - High, High, coincident with SI FU4e 1-out-of-2 Steam Line Isolation

- Steam Line Break

- Pressurizer Pressure - Low SI

- Containment Pressure - High SI

- Overpower T 1-out-of-2 in either loop SG Water Level - High, High FU5b 2-out-of-3 Feedwater Isolation

- Excessive Load Increase

- Reduction in feedwater enthalpy

- SI on Pressurizer Pressure -

Low

- SG Flow - High, High, coincident with SI

- SG Flow - High, coincident with SI and Low Tavg Provides backup (non-credited) feedwater isolation 2-out-of-3 in either loop

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 Page 27 of 27 Table E1-3(2) - Information to Support ESFAS Instrumentation Redundancy and Diversity ESFAS Function Functional Unit (FU)

Channels to Trip ESF Actuation Primary Design Basis Accident Diverse Actuation Function(s)

Comments SG Water Level - Low, Low FU6b 2-out-of-3 AFW Start

- Loss of Normal Feedwater

- Loss of all AC Auxiliaries

- Undervoltage Bus A01 and A02

- AMSAC 2-out-of-3 in either loop.

AMSAC inputs are separate and independent of RPS.

Undervoltage Bus A01 and A02 FU6d 1 of 2 taken twice AFW Start

- Loss of Normal Feedwater

- SG Water Level - Low, Low

- AMSAC Starts Turbine Driven AFW. Two channels of 4160V bus undervoltage relays are used to initiate AFW. Each channel contains two relays, which are combined in 1-out-of-2 coincidence.

Notes:

1.

2-out-of-3 logic also allows a channel in test to be tripped, while the remaining two channels provide protection in a 1-out-of-2 logic until the tested channel is restored to service

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Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Page 1 of 3 ENCLOSURE 3 Point Beach Units 1 and 2 License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b Information Supporting Technical Adequacy of PRA Models Without PRA Standards Endorsed by Regulatory Guide 1.200, Revision 2

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 3 Page 2 of 3 Table of Contents Section Title Page 1.0 Purpose............................................................................................................................. 3

Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 3 Page 3 of 3 1.0 Purpose This enclosure is not applicable to the Point Beach Nuclear Plant submittal. NextEra Energy Point Beach, LLC is not proposing to use any PRA models in its Risk-Informed Completion Time Program for which a PRA standard, endorsed by the NRC in RG 1.200, Revision 2 does not exist.

ML22140A142

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