ML22140A142
| ML22140A142 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 05/20/2022 |
| From: | Point Beach |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML22140A131 | List: |
| References | |
| NRC 2022-0007 | |
| Download: ML22140A142 (38) | |
Text
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 SSC PRA Design Success PRA Success Condition Applicable SSCs Comments Function(s) Modeled Criteria Criteria Description TS 3.3.1, RPS Instrumentation (Table 3.3-1)
The operator actions for failure to Condition B actuate a manual reactor trip will be One Manual 1 of 2 (FU1) Reactor Trip Not Modeled - See used as a surrogate to Reactor Trip No Manual Rx Trip Two Manual Rx Trip channels Initiation comments conservatively bound the risk channel inoperable channels increase associated with this (Modes 1,2) function as permitted by NEI 06-09.
2 of 4 The condition of one of two (FU2a)
Reactor Trip Power Range Not Modeled - See inoperable reactor trip breakers is Four Power Range Neutron Flux No Initiation Neutron Flux- High comments used as a surrogate for this High channels channels condition.
2 of 4 The condition of one of two (FU2b)
Reactor Trip Power Range Not Modeled - See inoperable reactor trip breakers is Four Power Range Neutron Flux No Initiation Neutron Flux- Low comments used as a surrogate for this Low channels channels condition.
The condition of one of two (FU5) 2 of 4 Reactor Trip Not Modeled - See inoperable reactor trip breakers is Four Overtemperature T No Overtemperature T Initiation comments used as a surrogate for this channels channels condition.
The condition of one of two 2 of 4 Condition D (FU6) Reactor Trip Not Modeled - See inoperable reactor trip breakers is No Overpower T One channel Four Overpower T channels Initiation comments used as a surrogate for this channels inoperable condition.
2 of 3 The condition of one of two (FU7b)
Reactor Trip Pressurizer Not Modeled - See inoperable reactor trip breakers is Three Pressurizer Pressure - No Initiation Pressure - High comments used as a surrogate for this High channels channels condition.
2 of 3 The condition of one of two (FU13)
Reactor Trip SG Water Level Not Modeled - See inoperable reactor trip breakers is Three SG Water Level Low-Low No Initiation Low-Low channels comments used as a surrogate for this channels per SG on any SG condition.
One SG Water Level (FU14) The condition of one of two Low coincident w/
Two SG Water Level Low; Reactor Trip Not Modeled - See inoperable reactor trip breakers is No one Steam-Flow/
Coincident w/ Steam-Flow/ Feed- Initiation comments used as a surrogate for this Feed-Flow Mismatch Flow Mismatch channels per SG condition.
channel on any SG
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 SSC PRA Design Success PRA Success Condition Applicable SSCs Comments Function(s) Modeled Criteria Criteria Description 1 of 2 The condition of one of two Condition E (FU12)
Reactor Trip Underfrequency Bus Not Modeled - See inoperable reactor trip breakers is One channel Two Underfrequency Bus A01, No Initiation channels on both comments used as a surrogate for this inoperable A02 channels per bus A01 and A02 condition.
2 of 4 The condition of one of two (FU7a)
Reactor Trip Pressurizer Not Modeled - See inoperable reactor trip breakers is Four Pressurizer Pressure - Low No Initiation Pressure - Low comments used as a surrogate for this channels channels condition.
2 of 3 The condition of one of two (FU8)
Reactor Trip Pressurizer Water Not Modeled - See inoperable reactor trip breakers is Three Pressurizer Water Level - No Initiation Level - High comments used as a surrogate for this High channels channels condition.
Condition K One channel 2 of 3 inoperable (FU9b) Reactor Coolant The condition of one of two Three Reactor Coolant Flow - Reactor Trip Flow - Low Not Modeled - See inoperable reactor trip breakers is No Low (Two loops) channels per Initiation channels on both comments used as a surrogate for this RCS Loop RCS Loops when condition.
above P8 1 of 2 The condition of one of two (FU11)
Reactor Trip Undervoltage Bus Not Modeled - See inoperable reactor trip breakers is Two Undervoltage Bus A01, A02 No Initiation channels on comments used as a surrogate for this channels per electrical bus A01 and A02 condition.
2 of 3 Condition L (FU9a) Reactor Coolant The condition of one of two One Reactor Three Reactor Coolant Flow - Reactor Trip Flow - Low Not Modeled - See inoperable reactor trip breakers is Coolant Flow-Low No Low (Single Loop) channels per Initiation channels on any comments used as a surrogate for this (Single Loop)
RCS Loop RCS Loop when condition.
channel inoperable below P8 Condition M One Reactor One RCP Breaker The condition of one of two (FU10a)
Coolant Pump Reactor Trip Position channel on Not Modeled - See inoperable reactor trip breakers is One RCP Breaker Position No Breaker Position Initiation any RCP when comments used as a surrogate for this (Single Loop) channel per RCP (Single Loop) below P8 condition.
channel inoperable One RCP breaker The condition of one of two Condition N (FU10b)
Reactor Trip Position channel on Not Modeled - See inoperable reactor trip breakers is One inoperable One RCP breaker position (two No Initiation both RCPs when comments used as a surrogate for this channel loops) channel per RCP above P8 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 SSC PRA Design Success PRA Success Condition Applicable SSCs Comments Function(s) Modeled Criteria Criteria Description 2 of 3 The condition of one of two (FU15a)
Reactor Trip Turbine trip - Low Not Modeled - See inoperable reactor trip breakers is Three Turbine trip - Low Auto- No Condition O Initiation Auto-Stop Oil comments used as a surrogate for this Stop Oil Pressure channels One turbine trip Pressure channels condition.
channel 2 of 3 The condition of one of two inoperable (FU15b)
Reactor Trip Turbine trip - Stop Not Modeled - See inoperable reactor trip breakers is Two Turbine trip - Stop Valve No Initiation Valve Closure comments used as a surrogate for this Closure channels channels condition.
The condition of one of two 1 of 2 (FU16) Reactor Trip Not Modeled - See inoperable reactor trip breakers is No SI input from ESFAS Condition P Two SI input from ESFAS trains Initiation comments used as a surrogate for this trains One train condition.
inoperable The condition of one of two 1 of 2 (Modes 1,2) (FU21) Reactor Trip Not Modeled - See inoperable reactor trip breakers is No Automatic Trip Logic Two Automatic Trip Logic trains Initiation comments used as a surrogate for this trains condition.
Condition Q This SSC is used as a surrogate for One RTB (FU18) Reactor Trip 1 of 2 Not Modeled - See No other TS 3.3.1 RPS Instrumentation inoperable Two RTB trains Initiation RTB trains comments Conditions.
(Modes 1,2)
Condition U One RTB (FU19) The condition of one of two One trip Undervoltage and One RTB Undervoltage and Reactor Trip Not Modeled - See inoperable reactor trip breakers is mechanism No Shunt Trip Shunt Trip Mechanism per RTB Initiation comments used as a surrogate for this inoperable for one Mechanism on any train condition.
RTB (Modes 1,2) RTB train TS 3.3.2, ESFAS Instrumentation (Table 3.3-2)
The operator actions for failure to Safety 1 of 2 manually actuate SI will be used as (FU1a) Not Modeled - See Injection No Manual Initiation a surrogate to conservatively bound Two Manual Initiation channels comments Condition B Initiation channels the risk increase associated with this One channel function as permitted by NEI 06-09.
inoperable The condition of manual SI function Containment 1 of 2 (FU3a) Not Modeled - See inoperable is used as a surrogate for Isolation No Manual Initiation Two Manual Initiation channels comments this condition since an SI signal Initiation channels 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 SSC PRA Design Success PRA Success Condition Applicable SSCs Comments Function(s) Modeled Criteria Criteria Description 1 of 2 The failure of the automatic SI (FU1b)
Safety Automatic signals will be used as a surrogate Two Automatic Not Modeled - See Injection No Actuation Logic and to conservatively bound the risk Actuation Logic and Actuation comments Initiation Actuation Relay increase associated with this Condition C Relay trains trains channels function as permitted by NEI 06-09.
One train 1 of 2 inoperable. (FU3b) The condition of automatic SI Containment Automatic Two Automatic Not Modeled - See function inoperable is used as a Isolation No Actuation Logic and Actuation Logic and Actuation comments surrogate for this condition since an Initiation Actuation Relay trains SI signal generates a CI signal.
Relay trains 2 of 3 The condition of automatic SI (FU1c) Safety Containment Not Modeled - See function inoperable is used as a Three Containment Pressure - Injection No Pressure - High comments surrogate for this condition since it is High channels Initiation channels the same function (SI initiation).
2 of 3 The condition of automatic SI (FU1d) Safety Pressurizer Not Modeled - See function inoperable is used as a Three Pressurizer Pressure - Injection No Pressure - Low comments surrogate for this condition since it is Low channels Initiation channels the same function (SI initiation).
2 of 3 The condition of automatic SI (FU1e) Safety Steam Line Not Modeled - See function inoperable is used as a Three Steam Line Pressure - Injection No Pressure - Low comments surrogate for this condition since it is Low channels per steam line Initiation channels per main Condition D the same function (SI initiation).
steam line One channel The failure of the model logic for inoperable. 2 of 3 steam generator isolation will be (FU4c) Steam Line Containment Not Modeled - See used as a surrogate to Three Containment Pressure - Isolation No Pressure - High, comments conservatively bound the risk High, High channels Initiation High channels increase associated with this function as permitted by NEI 06-09.
1 of 2 High Steam (FU4d) Flow channels Two High Steam Flow channels; Steam Line coincident with SI The condition of steam generator and coincident with 2 Not Modeled - See Coincident with SI and Isolation No isolation function inoperable is used Coincident with three Tavg - Low, of 3 Tavg - Low, Low comments Initiation as a surrogate for this condition.
Low channels per RCS loop channels
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 SSC PRA Design Success PRA Success Condition Applicable SSCs Comments Function(s) Modeled Criteria Criteria Description 1 of 2 (FU4e)
Steam Line High, High Steam The condition of steam generator Two High, High Steam Flow Not Modeled - See Isolation No Flow channels per isolation function inoperable is used channels per steam line; comments Initiation steam line coincident as a surrogate for this condition.
Coincident with SI with SI 2 of 3 (FU5b) Feedwater The condition of steam generator SG Water Level - Not Modeled - See Three SG Water Level - High Isolation No isolation function inoperable is used channels per SG High channels per comments Initiation as a surrogate for this condition.
SG The failure of the model logic for 2 of 3 (FU6b) Auxiliary these relays will be used as a SG Water Level - Not Modeled - See Three SG Water Level - Low, Feedwater No surrogate to conservatively bound Low channels per SG Low, Low channels comments Initiation the risk increase associated with this on any SG function as permitted by NEI 06-09.
Condition F (FU4a) Steam Line One Manual The condition of steam generator Not Modeled - See One channel One Manual Initiation channel Isolation No Initiation channel per isolation function inoperable is used comments inoperable per RCS loop Initiation RCS loop as a surrogate for this condition.
1 of 2 (FU4b) Steam Line The condition of steam generator Automatic Actuation Not Modeled - See Two Automatic Actuation Logic Isolation No isolation function inoperable is used Logic and Actuation comments and Actuation Relay trains Initiation as a surrogate for this condition.
Relay trains 1 of 2 (FU5a) The condition of AFW initiation is Condition G Feedwater Automatic Actuation Not Modeled - See Two Automatic Actuation Logic No used as a surrogate for this One train Isolation Logic and Actuation comments and Actuation Relay trains condition.
inoperable Relay trains The failure of the model logic for 1 of 2 (FU6a) these relays will be used as a Auxiliary Automatic Actuation Not Modeled - See Two Automatic Actuation Logic No surrogate to conservatively bound Feedwater Logic and Actuation comments and Actuation Relay trains the risk increase associated with this Relay trains function as permitted by NEI 06-09.
The failure of the model logic for 1 of 2 starting all four AFW pumps will be Condition H (FU6d)
Auxiliary Undervoltage Bus Not Modeled - See used as a surrogate to One channel Two Undervoltage Bus A01 and No Feedwater channels on comments conservatively bound the risk inoperable A02 channels per electrical bus A01 and A02 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 SSC PRA Design Success PRA Success Condition Applicable SSCs Comments Function(s) Modeled Criteria Criteria Description TS 3.4, Reactor Coolant System TS 3.4.11 Condition B RCS One PORV Two Power Operated Relief 1 of 2 pressure Yes Same inoperable and not Valves (PORVs) PORVs control capable of being manually cycled TS 3.4.11 Condition C RCS Associated Block Two PORV Block Valves Yes Same One block valve integrity Valve closure inoperable TS 3.5, Emergency Core Cooling System (ECCS) 1 of 2 SI pumps, and 1 of 2 RHR pumps Two ECCS trains each Emergency TS 3.5.2 comprised of one SI pump, one core cooling Condition A RHR pump, one RHR heat and post- 1 of 2 RHR pumps Yes w/ suction from Same One ECCS train exchanger and associated accident inoperable RWST and containment sump (long-term) containment sump, flowpaths core cooling supplying suction to 1 of 2 SI pumps for flowpath to RCS TS 3.6, Containment Systems TS 3.6.2 The failure of the model logic for Condition C One of two containment penetrations will be One or more One equipment hatch; One containment air lock Containment Not Modeled - See used as a surrogate to containment air personnel airlock; Two No doors closed with integrity comments conservatively bound the risk locks inoperable for emergency airlocks acceptable increase associated with this reasons other containment leakage function as permitted by NEI 06-09.
Condition A or B
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 SSC PRA Design Success PRA Success Condition Applicable SSCs Comments Function(s) Modeled Criteria Criteria Description TS 3.6.3 Condition A One or more penetration flow The failure of the model logic for paths with one containment penetrations will be containment 1 of 2 Two isolation valves on each Containment Not Modeled - See used as a surrogate to isolation valve No isolation valves per containment penetration integrity comments conservatively bound the risk inoperable penetration isolate increase associated with this (applicable to function as permitted by NEI 06-09.
penetration flow paths with two containment isolation valves)
TS 3.6.3 Condition C One or more penetration flow paths with one The failure of the model logic for containment containment penetrations will be One isolation valve and one Each isolation valve Containment Not Modeled - See used as a surrogate to closed system on each No isolation valve per inoperable integrity comments conservatively bound the risk containment penetration penetration isolates (applicable to increase associated with this penetration flow function as permitted by NEI 06-09.
paths with one containment isolation valve and a closed system)
TS 3.7, Plant Systems TS 3.7.2 Condition A Two Main Steam Lines equipped Steam Line One Steam with one Main Steam Isolation Isolation MSIV on affected Generator Yes Same Valve (MSIV) and one Non- from Faulted steam line isolates flowpath with one Return Check Valve Steam Line or more inoperable valves in MODE 1
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 SSC PRA Design Success PRA Success Condition Applicable SSCs Comments Function(s) Modeled Criteria Criteria Description Facilitate TS 3.7.4 Unit Condition A Two Atmospheric Dump Valves 1 of 2 cooldown to Yes Same One required ADV (ADVs) ADV flowpaths SDC flowpath inoperable conditions TS 3.7.5 Condition A Turbine driven AFW pump system Feedwater inoperable due to One turbine-driven AFW pump One motor driven supply to one inoperable and one motor-driven AFW pump AFW pump supplies SGs upon Yes Same steam supply, and associated CST and SW CST feedwater to loss of main OR Turbine driven suction piping both SGs feedwater AFW pump system inoperable in MODE 3 following refueling TS 3.7.5 Condition B Feedwater One turbine-driven AFW pump One turbine driven One AFW pump supply to and one motor-driven AFW pump AFW pump supplies system inoperable SGs upon Yes Same and associated CST and SW CST feedwater to in MODE 1, 2 or 3 loss of main suction piping both SGs for reasons other feedwater than Condition A Heat sink for Two Component Cooling Water removing process and 1 CCW pump and TS 3.7.7 (CCW) trains each consisting of operating 1 CCW HX provide Condition A one CCW pump and one CCW heat from Yes heat sink to Same One CC pump heat exchanger, and one safety- safety related inoperable common CCW heat exchanger related equipment capable of aligning to either train components
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 SSC PRA Design Success PRA Success Condition Applicable SSCs Comments Function(s) Modeled Criteria Criteria Description Heat sink for Two Component Cooling Water removing TS 3.7.7 process and 1 CCW pump and (CCW) trains each consisting of Condition B operating 1 CCW HX provide one CCW pump and one CCW One required CC heat from Yes heat sink to Same heat exchanger, and one heat exchanger safety- safety related common CCW heat exchanger inoperable related equipment capable of aligning to either train components Heat sink for 2 SW pumps and TS 3.7.8 removing Six Service Water (SW) system 1 SW Ring Header Condition A process and pumps, one common ring provide heat sink to One SW pump operating header, non-essential flowpath Yes CCW system and Same inoperable AND heat from isolation valves and associated essential loads; Both units in Modes safety-SW intake piping auto-isolate non-1, 2, 3, or 4 related essential flowpaths components Heat sink for 2 SW pumps and TS 3.7.8 removing Six Service Water (SW) system 1 SW Ring Header Condition C process and pumps, one common ring provide heat sink to SW ring header operating header, non-essential flowpath Yes CCW system and Same continuous heat from isolation valves and associated essential loads; flowpath safety-SW intake piping auto-isolates non-interrupted related essential flowpaths components TS 3.7.8 Condition D One or more Heat sink for 3 SW pumps and non-essential-SW- removing Six Service Water (SW) system 1 SW Ring Header load flowpath(s) process and pumps, one common ring provide heat sink to with one required operating header, non-essential flowpath Yes CCW system and Same automatic isolation heat from isolation valves and associated essential loads; valve inoperable. safety-SW intake piping auto-isolates non-AND Affected non- related essential flowpaths essential components flowpath(s) not isolated
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 SSC PRA Design Success PRA Success Condition Applicable SSCs Comments Function(s) Modeled Criteria Criteria Description TS 3.8, Electrical Power Systems Power onsite TS 3.8.1 safeguards Condition A buses Associated unit from offsite 345/13.8 kV Two 13.8 kV station auxiliary and onsite (X03) transformer transformers; one 13.8 kV gas transmission Automatically power inoperable.
turbine; two Class 1E 4.16 kV networks to Yes associated ESF Same OR electrical buses; four 4.16 kV support busses Gas turbine not in diesel generators (DG) normal, safe operation when shutdown utilizing opposite and accident units 345/13.8 kV mitigation (X03) transformer.
conditions Power onsite safeguards buses from offsite TS 3.8.1 Two 13.8 kV station auxiliary and onsite Condition B transformers; one 13.8 kV gas transmission Automatically power Associated unit's turbine; two Class 1E 4.16 kV networks to Yes associated ESF Same 13.8/4.16kV (X04) electrical buses; four 4.16 kV support busses transformer diesel generators (DG) normal, safe inoperable shutdown and accident mitigation conditions
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 SSC PRA Design Success PRA Success Condition Applicable SSCs Comments Function(s) Modeled Criteria Criteria Description Power onsite TS 3.8.1 safeguards Condition C buses Associated unit's from offsite required offsite Two 13.8 kV station auxiliary and onsite power source to transformers; one 13.8 kV gas transmission Automatically power buses A05 and A06 turbine; two Class 1E 4.16 kV networks to Yes associated ESF Same inoperable. electrical buses; four 4.16 kV support busses OR Required offsite diesel generators (DG) normal, safe power source to shutdown buses 1A05 and and accident 2A06 inoperable mitigation conditions Power onsite safeguards TS 3.8.1 buses Condition D from offsite One or more Two 13.8 kV station auxiliary and onsite required transformers; one 13.8 kV gas transmission Automatically power offsite power turbine; two Class 1E 4.16 kV networks to Yes associated ESF Same source(s) to one or electrical buses; four 4.16 kV support busses more required diesel generators (DG) normal, safe Class 1E 4.16 kV shutdown bus(es) and accident inoperable mitigation conditions TS 3.8.1 Power onsite Condition F safeguards One or more buses required Two 13.8 kV station auxiliary from offsite offsite power transformers; one 13.8 kV gas and onsite Automatically power source to one or turbine; two Class 1E 4.16 kV transmission Yes associated ESF Same more Class 1E 4.16 electrical buses; four 4.16 kV networks to busses kV safeguards diesel generators (DG) support bus(es) inoperable normal, safe AND Standby shutdown emergency power and accident inoperable to
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 SSC PRA Design Success PRA Success Condition Applicable SSCs Comments Function(s) Modeled Criteria Criteria Description redundant mitigation equipment conditions Ensure availability of TS 3.8.4 required DC Condition A 4 battery banks and associated power to Automatically power One DC electrical chargers and motor control shut down Yes associated ESF Same power subsystem centers (MCC) the reactor busses inoperable and maintain it in a safe condition Ensure availability of required DC TS 3.8.7 power to Two 120 VAC Condition A Four 120 VAC Instrument shut down Yes Instrument Inverters Same One required Inverters the reactor per electrical train inverter inoperable and maintain it in a safe condition
Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 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 Enclosure 1 Page 16 of 27 Table E1 In-Scope TS/LCO Conditions RICT Estimate RICT TS/LCO Required Action Condition Description 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 One Reactor Coolant Pump Breaker Position, (Single Loop) channel 3.3.1.M 30 inoperable 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 One or more, containment air locks inoperable for reasons other than 3.6.2.C 9 Condition A or B Containment Isolation Valves One or more penetration flow paths with one containment isolation valve 3.6.3.A 9 inoperable One or more penetration flow paths with one containment isolation valve 3.6.3.C 9 inoperable Main Steam Isolation Valves (MSIVs) and Non-Return Check Valves One Steam Generator, flowpath with one or more inoperable valves in 3.7.2.A 30 MODE 1 Atmospheric Dump Valve (ADV) Flowpaths 3.7.4.A One required ADV flowpath inoperable 30 Auxiliary Feedwater (AFW)
Turbine driven AFW pump system inoperable due to one inoperable steam 3.7.5.A supply, OR Turbine driven AFW pump system inoperable in MODE 30 3 following refueling One AFW pump system inoperable in MODE 1, 2 or 3 for reasons other than 3.7.5.B 30 Condition A
Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 Page 17 of 27 Table E1 In-Scope TS/LCO Conditions RICT Estimate RICT TS/LCO Required Action Condition Description 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 One or more non-essential-SW-load flowpath(s) with one required 3.7.8.D automatic isolation valve inoperable AND Affected non-essential flowpath(s), 30 not isolated.
AC Sources - Operating Associated unit 345/13.8 kV (X03) transformer inoperable OR Gas turbine 3.8.1.A 30 not in operation when utilizing opposite units 345/13.8 kV (X03) transformer.
3.8.1.B Associated unit's 13.8/4.16kV (X04) transformer, inoperable 30 Associated unit's required offsite power source to buses A05 and A06 3.8.1.C inoperable OR Required offsite power source to buses 1A05 and 2A06 30 inoperable.
One or more required offsite power source(s) to one or more required Class 3.8.1.D 30 1E 4.16 kV bus(es) inoperable.
One or more required offsite power source to one or more Class 1E 4.16 kV 3.8.1.F safeguards bus(es) inoperable AND Standby emergency power inoperable 8 to redundant equipment.
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 Enclosure 1 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.
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 Enclosure 1 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 Enclosure 1 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 Channels Functional Primary Transient /
RPS Function to Trip Diverse Actuation Functions() (1) Comments Unit (FU) (1)(2)(3) Accident Protection (4)
- Uncontrolled RCCA Withdrawal at Power
- Uncontrolled RCCA - Overtemperature T Withdrawal from a - Overpower T Power Range Neutron No interlocks (i.e. always FU2a 2 out of 4 Subcritical Condition - Pressurizer Pressure - High Flux (high setting) active)
- Steam Line Break - Pressurizer Level - High
- Loss of External - Manual Trip Electrical Load
- Loss of RCS Coolant
- Uncontrolled RCCA - Source Range Flux Manual block above P10 Withdrawal from a - Intermediate Range Flux interlock (~10% RTP)
Power Range Neutron FU2b 2 out of 4 Subcritical Condition - Power Range Neutron Flux - High allowed. Automatically Flux (low setting)
- Loss of RCS Coolant - Manual Trip reinstated below P10.
- Uncontrolled RCCA Withdrawal from a Subcritical Condition
- Uncontrolled RCCA
- Power Range Neutron Flux - High Withdrawal at Power
- Overpower T RCCA Drop No interlocks. Requires two
- Pressurizer Pressure - High
- Loss of RCS Flow sets of temperature Overtemperature T FU5 2 out of 4 - Pressurizer Level - High
- RCCA Drop measurements above trip
- SG Level Low-Low
- Loss of External setpoint per RCS loop
- Turbine Trip Electrical Load
- Loss of Normal Feedwater
- SG Tube Rupture
- Loss of RCS Coolant
- Uncontrolled RCCA No interlocks. Requires two Withdrawal at Power
- Source Range Neutron Flux sets of temperature Overpower T FU6 2 out of 4 - RCCA Drop
- Manual Trip measurements above trip
- Steam Line Break setpoint per RCS loop
- CVCS Malfunction
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 Channels Functional Primary Transient /
RPS Function to Trip Diverse Actuation Functions() (1) Comments Unit (FU) (1)(2)(3) Accident Protection (4)
- Uncontrolled RCCA
- Power Range Neutron Flux - High Withdrawal at Power
- Overtemperature T
- Loss of External Pressurizer Pressure - - Pressurizer Level - High FU7a 2 out of 3 Electrical Load No interlocks High - SG Level Low-Low
- Loss of RCS Flow
- Loss of Normal
- RCCA Drop - Overtemperature T Pressurizer Pressure - - SG Tube Rupture - Overpower T Automatic block below P-7 FU7b 2 out of 4 Low - Steam Line Break - Safety Injection Reinstated above P-7
- Loss of RCS Coolant - Manual Trip
- Uncontrolled RCCA Withdrawal at Power - Pressurizer Pressure - High
- Loss of External - Power Range Neutron Flux - High Pressurizer Level - High FU8 2 out of 3 Automatic block below P-7 Electrical Load - Overtemperature T
- Loss of Normal - Manual Trip Feedwater One loop trip requires 2 of 3 channels on either RCS loop.
- Overtemperature T Blocked below P8 interlock
- Pressurizer Pressure - High
- Loss of RCS Flow (~35% RTP). Reinstated RCS Flow - Low FU9a - Undervoltage Bus A01, A02 2 out of 3 - Loss of All AC Power above P8.
(one loop) (two loops) FU9b - Underfrequency Bus A01, A02 to the Auxiliaries Two loop trip requires 2 of 3
- RCP Breaker position channels on both RCS loops.
- Manual Trip Blocked below P7 (~10%
RTP). Reinstated above P7.
- Overtemperature T No interlocks. ~5 second RCP breaker position -
FU10a - Loss of RCS Flow - Pressurizer Pressure - High time delay. Trips RCP on one loop 1 out of 1 FU10b - Loss of All AC Power - Undervoltage Bus A01, A02 affected bus. One loop RCP breaker position - per loop to the Auxiliaries - RCS Flow - Low (one loop) (two loops) interlocked with P-8. Two 2 loops
- Manual Trip loops interlocked with P-7.
- Overtemperature T
- Loss of RCS Flow - Pressurizer Pressure - High Undervoltage Bus A01, 1 out of 2 FU11 - Loss of All AC Power - RCP Breaker position Automatic block below P-7 A02 per bus to the Auxiliaries - RCS Flow - Low (one loop) (two loops)
- Manual Trip
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 Channels Functional Primary Transient /
RPS Function to Trip Diverse Actuation Functions() (1) Comments Unit (FU) (1)(2)(3) Accident Protection (4)
- Loss of RCS Flow Underfrequency Bus A01, 1 out of 1 - Open RCP Breaker position No interlocks. Trips RCP on FU12 - Loss of All AC Power A02 per bus - Manual Trip affected bus.
to the Auxiliaries
- Steam/Feed flow mismatch coincident w/ low SG level
- Overtemperature T
- Loss of Normal
- Overpressure T Feedwater
- Pressurizer Pressure - High SG Water Level - Low, - Loss of External Requires 2 of 3 low-low level FU13 2 out of 3 - Pressurizer Level - High Low Electrical Load channels in either SG
- RCS Flow - Low
- Loss of All AC Power
- Undervoltage Bus A01, A02 to the Auxiliaries
- Underfrequency Bus A01, A02
- SG Water Level - Low, Low
- Overtemperature T
- Overpressure T
- Loss of Normal - Pressurizer Pressure - High SG Water Level Low; 1 out of 2 Feedwater - Pressurizer Level - High *Coincident w/1 out of 2 low coincident w/ Steam Flow FU14 per loop* - Loss of All AC Power - RCS Flow - Low SG water level per loop
/ Feed Flow Mismatch to the Auxiliaries - Undervoltage Bus A01, A02
- Underfrequency Bus A01, A02
- SG Water Level - Low, Low
- Loss of Normal - Pressurizer Pressure - High Feedwater - Overtemperature T Turbine trip on low oil FU15a 2 out of 3 - Loss of External - Overpower T Interlocked with P-7 and P-9 pressure Electrical Load - Undervoltage Bus A01, A02
- Underfrequency Bus A01, A02
- Manual Trip
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 Channels Functional Primary Transient /
RPS Function to Trip Diverse Actuation Functions() (1) Comments Unit (FU) (1)(2)(3) Accident Protection (4)
- SG Water Level - Low, Low
- Loss of Normal
- Pressurizer Pressure - High Feedwater
- Overtemperature T Turbine trip on stop valve - Loss of External FU15b 2 out of 2 - Overpower T Interlocked with P-7 and P-9 closure Electrical Load
- Undervoltage Bus A01, A02
- Underfrequency Bus A01, A02
- Manual Trip 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.
Point Beach Nuclear Plant, Units 1 and 2 NRC 2022-0007 Docket Nos. 50-266 and 50-301 Enclosure 1 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 Functional Channels ESF Primary Design Diverse Actuation ESFAS Function Comments Unit (FU) to Trip Actuation Basis Accident Function(s)
Containment Pressure - 2-out-of-3 Safety - Large Break LOCA FU1c - Pressurizer Pressure - Low SI High (1) Injection - Small Break LOCA
- Large Break LOCA - Containment Pressure - High Pressurizer Pressure - Safety - Small Break LOCA SI FU1d 2-out-of-3 Low Injection
- SG Tube Rupture - SG Water Level - High, High Safety SG Pressure - Low, Low FU1e 2-out-of-3 - Steam Line Break - Pressurizer Pressure - Low SI 2-out-of-3 in either loop Injection
- Small Break LOCA - Pressurizer Pressure - Low SI Containment Pressure - Steam Line - Large Break LOCA - Containment Pressure - High FU4c 2-out-of-3 High, High Isolation SI
- Steam Line Break - Overpower T 1-out-of-2 in either loop and 2-out-of-4 low Tavg channels. Four steam
- Pressurizer Pressure - Low SI flow channels are divided SG Flow - High, 1-out-of-2 Steam Line - Containment Pressure - High into two pairs, with one coincident with SI FU4d - Steam Line Break taken twice Isolation SI pair assigned to each and Low Tavg
- Overpower T steam loop. Flow signals from each loop are combined in a 1-out-of-2 coincidence logic.
- Pressurizer Pressure - Low SI SG Flow - High, High, Steam Line - Containment Pressure - High FU4e 1-out-of-2 - Steam Line Break 1-out-of-2 in either loop coincident with SI Isolation SI
- Overpower T
- SI on Pressurizer Pressure -
- Excessive Load Provides backup (non-Low SG Water Level - High, Feedwater Increase - SG Flow - High, High, credited) feedwater FU5b 2-out-of-3 isolation High Isolation coincident with SI
- Reduction in - SG Flow - High, coincident feedwater enthalpy 2-out-of-3 in either loop with SI and Low Tavg
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 Functional Channels ESF Primary Design Diverse Actuation ESFAS Function Comments Unit (FU) to Trip Actuation Basis Accident Function(s)
- Loss of Normal 2-out-of-3 in either loop.
Feedwater - Undervoltage Bus A01 and SG Water Level - Low, AMSAC inputs are FU6b 2-out-of-3 AFW Start A02 Low separate and
- Loss of all AC - AMSAC independent of RPS.
Auxiliaries Starts Turbine Driven AFW. Two channels of 4160V bus undervoltage Undervoltage Bus 1 of 2 - Loss of Normal - SG Water Level - Low, Low relays are used to initiate FU6d AFW Start A01 and A02 taken twice Feedwater - AMSAC 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.