Text

Duke Energy - Application to Revise the Technical Specification for Engineered Safety Feature Actuation System (ESFAS) Instrumentation
	ML20338A264
Person / Time
Site:	Mcguire, Catawba, Harris, McGuire ; (NPF-009, NPF-017, NPF-035, NPF-052, NPF-063)
Issue date:	12/03/2020
From:	Snider S; Duke Energy Carolinas, Duke Energy Progress
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RA-20-0207
	Download: ML20338A264 (43)
	v • d • e

( ~ DUKE ENERGY RA-20-0207 December 3, 2020 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 CATAWBA NUCLEAR STATION, UNITS 1 AND 2 Steve Snider Vice President Nuclear Engineering 526 South Church Street, EC-07H Charlotte, NC 28202 980-373-6195 Steve.Snider@duke-energy.com 10 CFR 50.90 DOCKET NOS. 50-413 AND 50-414 / RENEWED LICENSE NOS. NPF-35 AND NPF-52 MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-369 AND 50-370 / RENEWED LICENSE NOS. NPF-9 AND NPF-17 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400 / RENEWED LICENSE NO. NPF-63

SUBJECT:

APPLICATION TO REVISE THE TECHNICAL SPECIFICATION FOR ENGINEERED SAFETY FEATURE ACTUATION SYSTEM (ESFAS)

INSTRUMENTATION Pursuant to 10 CFR 50.90, Duke Energy Carolinas, LLC, and Duke Energy Progress, LLC, collectively referred to henceforth as "Duke Energy," is submitting a request for amendments to the Technical Specifications (TS) for Catawba Nuclear Station (CNS), Units 1 and 2; McGuire Nuclear Station (MNS), Units 1 and 2; and Shearon Harris Nuclear Power Plant (HNP), Unit 1.

For CNS and MNS, the proposed change would revise Table 3.3.2-1, Function 8.a. (ESFAS Interlocks, Reactor Trip, P-4) of TS 3.3.2, "Engineered Safety Feature Actuation System (ESFAS) Instrumentation." For HNP, the proposed change would revise Table 3.3-3, Functional Unit 1 0.c. (Engineered Safety Features Actuation System Interlocks, Reactor Trip, P-4) of TS 3/4.3.2, "Engineered Safety Features Actuation System Instrumentation." Duke Energy is proposing to add a footnote to the applicable modes portion of the P-4 interlock Function to identify the enabled functions (i.e., sub-functions of the interlock required to meet the Limiting Condition for Operation) and the applicable MODES for each enabled function. In doing so, Duke Energy is proposing the removal of the turbine trip function of the P-4 interlock in MODE 3 from the existing specifications. For CNS only, Duke Energy is proposing the removal of the steam dump function of the P-4 interlock in MODES 1, 2, and 3.

The Enclosure provides a description and assessment of the proposed change. Attachment 1 provides the existing CNS, MNS and HNP TS pages marked up to show the proposed change. provides existing CNS, MNS and HNP TS Bases pages marked up to show the proposed change. Changes to the existing CNS, MNS and HNP TS Bases will be implemented under the Technical Specification Bases Control Program for each station. The changes to the TS Bases are provided in Attachment 2 for information only. The retyped TS pages will be provided to the NRC immediately prior to issuance of the approved amendments.

U.S. Nuclear Regulatory Commission RA-20-0207 Page2 The proposed change has been evaluated in accordance with 10 CFR 50.91 (a)(1) using criteria in 10 CFR 50.92(c), and it has been determined that the proposed change involves no significant hazards consideration. The bases for these determinations are included in.

This submittal contains no regulatory commitments.

Duke Energy requests approval of the proposed change within one year of the date the application is accepted by the NRC for review. Once approved, the license amendments will be implemented within 120 days.

In accordance with 10 CFR 50.91, a copy of this application, with Attachments, is being provided to the designated North Carolina and South Carolina officials.

If you should have any questions regarding this submittal, or require additional information, please contact Art Zaremba, Manager - Nuclear Fleet Licensing, at 980-373-2062.

I declare under penalty of perjury that the foregoing is true and correct. Executed on December 3, 2020.

Sincerely, Steve Snider Vice President - Nuclear Engineering JLV

Enclosure:

Description and Assessment of the Proposed Change Attachments: 1. Proposed Technical Specifications Changes (Mark-up)

2. Proposed Technical Specifications Bases Changes (Mark-up)- For Information Only

U.S. Nuclear Regulatory Commission RA-20-0207 Page 3 cc (with Enclosure and Attachments):

L. Dudes, USNRC Region II - Regional Administrator J. Austin, USN RC Senior Resident Inspector - CNS A Hutto, USNRC Senior Resident Inspector-MNS J. Zeiler, USN RC Senior Resident Inspector - HNP M. Mahoney, NRR Project Manager-HNP J. Klos, NRR Project Manager - MNS K. Cotton, NRR Project Manager - CNS W. L. Cox, 111, Section Chief, North Carolina Department of Health and Human Services, RP Section (NC)

L. Garner, Manager, Radioactive and Infectious Waste Management (SC)

Enclosure to RA-20-0207 Page 1 of 22 Enclosure Description and Assessment of the Proposed Change

Subject:

Application to Revise the Technical Specification for Engineered Safety Feature Actuation System (ESFAS) Instrumentation 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 System Design and Operation 2.2 Current Technical Specifications Requirements 2.3 Reason for the Proposed Change 2.4 Description of the Proposed Change

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination Analysis 4.4 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

ATTACHMENTS:

1. Proposed Technical Specifications Changes (Mark-up)

2. Proposed Technical Specifications Bases Changes (Mark-up)- For Information Only

Enclosure to RA-20-0207 Page 2 of 22 1.0

SUMMARY

DESCRIPTION For Catawba Nuclear Station (CNS), Units 1 and 2, and McGuire Nuclear Station (MNS), Units 1 and 2, the proposed change would revise Table 3.3.2-1, Function 8.a. (ESFAS Interlocks, Reactor Trip, P-4) of TS 3.3.2, "Engineered Safety Feature Actuation System (ESFAS)

Instrumentation." For Shearon Harris Nuclear Power Plant (HNP), Unit 1, the proposed change would revise Table 3.3-3, Functional Unit 1 0.c. (Engineered Safety Features Actuation System Interlocks, Reactor Trip, P-4) and Table 4.3-2, Channel Functional Unit 1 0.c. (Engineered Safety Features Actuation System Interlocks, Reactor Trip, P-4) of TS 3/4.3.2, "Engineered Safety Features Actuation System Instrumentation." For CNS, MNS and HNP, Duke Energy is proposing to add a footnote to the applicable modes portion of the P-4 interlock Function to identify the enabled functions (i.e., sub-functions of the interlock required to meet the Limiting Condition for Operation) and the applicable MODES for each enabled function. In doing so, Duke Energy is proposing the removal of the turbine trip function of the P-4 interlock in MODE 3 from the existing specifications. For CNS only, Duke Energy is proposing the removal of the steam dump function of the P-4 interlock in MODES 1, 2, and 3. The P-4 function related to the steam dump function is not discussed at all in the HNP and MNS TS and TS Bases.

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation The following system design and operation discussion is applicable to CNS, MNS and HNP unless otherwise noted.

The ES FAS initiates necessary safety systems, based on the values of selected unit parameters, to protect against violating core design limits and the Reactor Coolant System (RCS) pressure boundary, and to mitigate accidents. The ESFAS is segmented into three distinct but separate modules:

field transmitters or process sensors and instrumentation; signal processing equipment including analog protection system, field contacts, and protection channel sets; and Solid State Protection System (SSPS) including input, logic, and output bays.

For this amendment request, the focus is on the SSPS segment of the ESFAS. The SSPS initiates the proper unit shutdown or engineered safety feature (ESF) actuation in accordance with the defined logic and based on the bistable outputs from the signal process control and protection system. The SSPS equipment is used for the decision logic processing of outputs from the signal processing equipment bistables. To meet the redundancy requirements, two trains of SSPS, each performing the same functions, are provided. If one train is taken out of service for maintenance or test purposes, the second train will provide ESF actuation for the unit. If both trains are taken out of service or placed in test, a reactor trip will result.

The SSPS performs the decision logic for most ESF equipment actuation; generates the electrical output signals that initiate the required actuation; and provide the status, permissive, and annunciator output signals to the control room. The bistable outputs from the signal processing equipment are sensed by the SSPS equipment and combined into logic matrices that represent combinations indicative of various transients. If a required logic matrix combination is completed, the system will send actuation signals via master and slave relays to those components whose aggregate function best serves to alleviate the condition and restore the unit to a safe condition.

Enclosure to RA-20-0207 Page 3 of 22 The required channels of ESFAS instrumentation provide unit protection in the event of any of the analyzed accidents. However, to allow some flexibility in unit operations, several interlocks are included as part of the ESFAS. These interlocks permit the operator to block some signals, automatically enable other signals, prevent some actions from occurring, and cause other actions to occur. The interlock Functions in Technical Specifications back up manual actions to ensure bypassable functions are in operation under the conditions assumed in the safety analyses. One of these interlocks, and the subject of this amendment request, is the Reactor Trip, P-4 ESFAS interlock. The P-4 interlock is enabled when a reactor trip breaker and its associated bypass breaker is open. The functions of the P-4 interlock, with respect to the Technical Specifications, are the following:

Trip the main turbine; Isolate Main Feedwater (MFW) with coincident low Tav9; Prevent reactuation of Safety Injection (SI) after a manual reset of SI; Prevent opening of the MFW isolation valves if they were closed on SI or High Steam Generator Water Level signal; and Transfer the steam dump from the load rejection controller to the unit trip controller.

(Note: CNS mentions this enabling function of the P-4 interlock function in the Technical Specifications Bases and MNS/HNP do not.)

2.2 Current Technical Specifications Requirements Catawba Technical Specification (TS) 3.3.2, "Engineered Safety Feature Actuation System (ESFAS)

Instrumentation," Table 3.3.2-1, contains the Function 8.a. for ESFAS Interlocks, Reactor Trip, P-4. The applicable MODES for Function 8.a. are 1, 2, and 3, without exception.

The CNS TS 3.3.2 Bases state that the functions of the P-4 interlock are:

Trip the main turbine; Isolate MFWwith coincident low Tav9; Prevent reactuation of SI after a manual reset of SI; Transfer the steam dump from load rejection controller to the unit trip controller; and Prevent opening of the MFW isolation valves if they were closed on SI or SG Water Level-High High.

The CNS TS define the modes of operation as MODE 1 - Power Operation, MODE 2 - Startup, MODE 3 - Hot Standby [average reactor coolant system (RCS) temperature;:: 350 degrees Fahrenheit (°F)], MODE 4 - Hot Shutdown [average RCS temperature < 350 °F and > 200 °F],

and MODE 5 - Cold Shutdown [average RCS temperature s 200 °F].

McGuire TS 3.3.2, "Engineered Safety Feature Actuation System (ESFAS) Instrumentation,"

Table 3.3.2-1, contains the Function 8.a. for ESFAS Interlocks, Reactor Trip, P-4. The applicable MODES for Function 8.a. are 1, 2, and 3, without exception.

Enclosure to RA-20-0207 Page 4 of 22 The MNS TS 3.3.2 Bases state that the functions of the P-4 interlock are:

Trip the main turbine; Isolate MFW coincident low Tav9; Prevent reactuation of SI after a manual reset of SI; and Prevent opening of the MFW isolation valves if they were closed on SI or SG Water Level-High High.

The MNS TS define the modes of operation as MODE 1 - Power Operation, MODE 2 - Startup, MODE 3 - Hot Standby [average reactor coolant system (RCS) temperature ;:: 350 degrees Fahrenheit (°F)], MODE 4 - Hot Shutdown [average RCS temperature < 350 °F and > 200 °F],

and MODE 5 - Cold Shutdown [average RCS temperature s 200 °F].

TS 3/4.3.2, "Engineered Safety Features Actuation System Instrumentation," Table 3.3-3, contains the Functional Unit 1 0.c. for Engineered Safety Features Actuation System Interlocks, Reactor Trip, P-4. The applicable MODES for Functional Unit 1 0.c. are 1, 2, and 3, without exception. Table 4.3-2 for ESFAS instrumentation surveillance requirements also contains Channel Functional Unit 1 0.c. for the P-4 interlock and the MODES for which the surveillance is required are 1, 2, and 3, without exception.

The HNP TS 3/4.3 Bases (Instrumentation) state that the ESFAS interlocks perform the following functions for the P-4 interlock:

Reactor Tripped - Actuates turbine trip, closes main feedwater valves on T avg, prevents the opening of the main feedwater valves which were closed by a Safety Injection or High Steam Generator Water Level signal, allows Safety Injection block so that components can be reset or tripped.

The HNP TS define the modes of operation as MODE 1 - Power Operation, MODE 2 - Startup, MODE 3 - Hot Standby [average reactor coolant system (RCS) temperature ;:: 350 degrees Fahrenheit (°F)], MODE 4 - Hot Shutdown [average RCS temperature < 350 °F and > 200 °F],

and MODE 5 - Cold Shutdown [average RCS temperature s 200 °F].

2.3 Reason for the Proposed Change Duke Energy is proposing changes to the ESFAS Instrumentation TS for CNS, MNS and HNP to prevent unnecessary trips of the main turbine during turbine warm up that adversely impacts startup and shutdown evolutions. In MODE 3, surveillance testing activities such as rod drop testing, utilize the closing and opening of reactor trip breakers. The block of a turbine trip on a reactor trip is only applied presently in MODES 4 or below. The proposed change would allow the block of the turbine trip on a reactor trip in MODE 3.

2.4 Description of the Proposed Change Catawba New footnote "(h)" is proposed to be added to the "APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS" portion of Function 8.a. of TS 3.3.2, Table 3.3.2-1, to identify the

Enclosure to RA-20-0207 Page 5 of 22 enabled functions and the applicable MODES for the P-4 interlock function. The new footnote states:

(h)

The functions of the Reactor Trip, P-4 interlock required to meet the LCO are:

Trip the main turbine - MODES 1 and 2 Isolate M FW with coincident low T avg - MODES 1, 2, and 3 Prevent reactuation of SI after a manual reset of SI - MODES 1, 2, and 3 Prevent opening of MFIVs if closed on SI or SG Water Level - High High -

MODES 1, 2, and 3 McGuire New footnote "(f)" is proposed to be added to the "APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS" portion of Function 8.a. of TS 3.3.2, Table 3.3.2-1, to identify the enabled functions and applicable MODES for the P-4 interlock function. The new footnote states:

(f)

The functions of the Reactor Trip, P-4 interlock required to meet the LCO are:

Trip the main turbine - MODES 1 and 2 Isolate M FW with coincident low T avg - MODES 1, 2, and 3 Prevent reactuation of SI after a manual reset of SI - MODES 1, 2, and 3 Prevent opening of MFIVs if closed on SI or SG Water Level - High High -

MODES 1, 2, and 3 New footnote "ti#-" is proposed to be added to the "APPLICABLE MODES" portion of Functional Unit 1 0.c. of TS 3/4.3.2, Table 3.3-3, to identify the enabled functions and applicable MODES for the P-4 interlock function. The new footnote states:

'11#-The functions of the Reactor Trip, P-4 interlock required to meet the LCO are:

Trip the main turbine - MODES 1 and 2 Isolate Main Feedwater with coincident low T avg - MODES 1, 2, and 3 Prevent reactuation of Safety Injection after a manual reset of Safety Injection -

MODES 1, 2, and 3 Prevent opening of Main Feedwater valves if closed on Safety Injection or Steam Generator Water Level - High High - MODES 1, 2, and 3 Additionally, new footnote "ti#-" is proposed to be added to the "MODES FOR WHICH SURVEILLANCE IS REQUIRED" portion of Channel Functional Unit 1 0.c. of TS 3/4.3.2, Table 4.3-2, to identify the enabled functions and applicable MODES for the P-4 interlock function.

The new footnote states:

Enclosure to RA-20-0207 Page 6 of 22 The functions of the Reactor Trip, P-4 interlock required to meet the LCO are:

Trip the main turbine - MODES 1 and 2 Isolate Main Feedwater with coincident low T avg - MODES 1, 2, and 3 Prevent reactuation of Safety Injection after a manual reset of Safety Injection

- MODES 1, 2, and 3 Prevent opening of Main Feedwater valves if closed on Safety Injection or Steam Generator Water Level - High High - MODES 1, 2, and 3

3.0 TECHNICAL EVALUATION

P-4 is a protection system interlock associated with the Reactor Protection System (RPS) and Engineered Safety Features Actuation System (ESFAS). The P-4 permissive is actuated when a reactor trip breaker and its bypass breaker are both opened. The control function addressed is that the P-4 actuates a turbine trip. This is done through direct feed contact to the turbine trip system. Thus, the presence of a P-4 interlock actuates a turbine trip to prevent excessive plant cooldown (positive reactivity addition) following a reactor trip. The P-4 interlock is currently required in MODES 1, 2, and 3. The purpose of this technical evaluation is to determine the consequences of removing the P-4 turbine trip function in MODE 3. Plant configurations during MODES 1, 2, and 3 are essential to evaluating the effects of a turbine trip and its removal from MODE 3. The turbine is brought online during MODE 1. The P-4 interlock turbine trip is needed to trip the turbine on reactor trip. During unit startup, in MODES 2 and 3, a small steam flow to the turbine can be used for turbine shell warming purposes and is considered to be negligible.

Until the turbine is placed online, secondary heat removal is accomplished by a combination of steam flow through steam dumps and steam generator blowdown to the condenser. During unit shutdown in MODE 3 the turbine is in a tripped condition. In general, the P-4 interlock function to trip the turbine on reactor trip is not required when in MODES 2 and 3 since the turbine is essentially in an "equivalent" tripped condition during a unit startup or an actual tripped condition during a unit shutdown. This justification is applicable to all sections documented below. The proposed change is to remove the TS requirement for the P-4 interlock turbine trip function in MODE 3.

3.1 Main Turbine Trip Updated Final Safety Analysis Report (UFSAR) Section 7.2.1.1.1, "Functional Performance Requirements" for CNS, MNS and HNP all describe the Reactor Protection System initiating a turbine trip signal whenever a reactor trip occurs in order to prevent insertion of positive reactivity, which could result from an overcooling of the RCS. In MODE 1 (Power Operation),

the main turbine is in service. Therefore, the P-4 interlock would be required to be in service during MODE 1 to provide a turbine trip upon reactor trip. However, in MODE 3 (Hot Standby),

the main turbine is on a turning gear and not in service. Steam flow in MODE 3 is aligned to steam dumps. The steam flow entering the turbine for warming purposes during MODE 3 can be small and is considered negligible. Therefore, MODE 3 would not be an applicable mode for the turbine trip function of the P-4 interlock. Hence, the only applicable mode for the turbine trip function of the P-4 interlock would be MODE 1.

There are two accident analyses performed in MODE 3 - UFSAR Section 15.4.1 Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low Power Startup Condition and UFSAR Section 15.4.6 CVCS Malfunction That Results in a Decrease in the Boron Concentration in the Reactor

Enclosure to RA-20-0207 Page 7 of 22 Coolant. The turbine trip on a reactor trip function in MODE 3 is not credited in the accident analysis to mitigate the consequences of these accidents.

A review of the UFSAR accident analyses for CNS, MNS and HNP determined that there are some non-loss-of-coolant accident (LOCA) transients during operation in MODES 1 and 2 that use the P-4 function of providing a turbine trip after a reactor trip. However, the proposed change does not impact the functionality of P-4 in MODES 1 and 2 (i.e., the turbine trip function will be maintained applicable in MODES 1 and 2). Other safety analyses transients either do not rely solely on the P-4 turbine trip or do not model the turbine trip on reactor trip function.

Further discussion of the accident analyses with respect to the turbine trip function of the P-4 interlock is provided in the following sections.

The events in Section 15.1 of the respective UFSARs address events that result in an increase in heat removal by the secondary system. These events are discussed individually below in Sections 3.1.1 through 3.1.4.

3.1.1 Feedwater System Malfunctions Feedwater system malfunctions are discussed in Section 15.1.1, "Feedwater System Malfunctions that Result in a Reduction in Feedwater Temperature" and Section 15.1.2, "Feedwater System Malfunctions Causing an Increase in Feedwater Flow" of the CNS and MNS UFSARs for a decreased feedwater temperature event and an increased feedwater flow event, respectively. The feedwater system malfunctions are discussed in these same UFSAR sections for HNP, except that the titles are slightly different for Section 15.1.1, "Feedwater System Malfunctions that Result in a Decrease in Feedwater Temperature" and Section 15.1.2 is "Feedwater System Malfunctions that Result in an Increase in Feedwater Flow." These events decrease the RCS temperature, which, in the presence of a negative moderator coefficient of reactivity, causes power to increase.

A reduction in feedwater temperature analyzed in Section 15.1.1 may be caused by a failure of a bypass valve that diverts flow around a portion of the feedwater heaters. This event is limiting at power (MODE 1) because the increased subcooling from the sudden reduction in feedwater inlet temperature to the steam generators creates a greater load demand on the RCS.

Nevertheless, for all 3 plants, the reduction in feedwater temperature is bounded by UFSAR Section 15.1.2 and/or 15.1.3 and no quantitative analysis is presented. Since the UFSAR Section 15.1.1 event is more limiting in MODE 1 and since it is bounded by the UFSAR Section 15.1.2 and/or Section 15.1.3 event, it is not impacted by the proposed change to remove the turbine trip function of the P-4 interlock in MODE 3. The feedwater system malfunction that results in a reduction in feedwater temperature presented in CNS, MNS, HNP UFSAR Section 15.1.1 remains bounded by the transient in Section 15.1.2 and/or Section 15.1.3 when the P-4 turbine trip on reactor trip function is removed.

Excessive feedwater flow could be caused by a full opening of a feedwater control valve due to a feedwater control system malfunction. In Section 15.1.2 of the CNS, MNS and HNP UFSARs, this event is analyzed at power (MODE 1) and zero power (MODE 2). The MODE 1 and MODE 2 analysis bound MODE 3 and below because a greater required shutdown margin (SDM) is present in MODE 3 and enough positive reactivity must be introduced to overcome the initial SDM before any return to power can occur. The MODE 3 analysis would be bounded by the MODE 2 analysis which starts from a critical condition which will result in a higher return to power than the MODE 3 analysis. Furthermore, in MODE 3 the turbine is offline and essentially

Enclosure to RA-20-0207 Page 8 of 22 in a tripped condition. The existing MODE 1 and MODE 2 analyses presented in Section 15.1.2 of the respective UFSARs remain bounding for the P-4 interlock proposed change.

3.1.2 Excessive Load Increase The excessive load increase analysis is presented in Section 15.1.3, "Excessive Increase in Secondary Steam Flow," of the CNS, MNS and HNP UFSARs. An excessive load increase event could be caused by an equipment malfunction in the steam dump control or turbine speed control. This event decreases the RCS (i.e., moderator) temperature, which causes power to increase in the presence of a negative moderator coefficient of reactivity. For CNS, MNS and HNP, the RCS is designed to accommodate a 10 percent step load increase or a 5 percent per minute ramp load increase in the range of 15 to 100 percent of full power, without actuating the Reactor Protection System. Increases in steam flow of more than 10 percent are addressed by the steam line break (SLB) events in Section 15.1.5, "Steam System Piping Failure" of the CNS, MNS and HNP UFSARs. In MODE 3, since the turbine is on the turning gear, the steam demand to the turbine is near zero, with minimal steam being supplied to warm up in preparation of turbine loading in MODE 1. Therefore, the SLB analysis bounds an excessive load increase event initiating from MODE 3. Consequently, an excessive load increase event occurring in MODE 3 and below is not impacted by the proposed change to remove the turbine trip function of the P-4 interlock in MODE 3.

3.1.3 Inadvertent Opening of a Steam Generator Relief or Safety Valve The steam release because of an inadvertent opening of a steam generator (SG) relief or safety valve results in an initial increase in steam flow which decreases during the accident as the steam pressure falls. The energy removal from the RCS causes a reduction in coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in an insertion of positive reactivity.

For CNS and MNS, UFSAR Section 15.1.4, "Inadvertent Opening of a Steam Generator Relief or Safety Valve," states that the analyses performed in Section 15.1.5, "Steam System Piping Failure" bounds the inadvertent opening of a SG relief or safety valve. For HNP, UFSAR Section 15.1.4, "Inadvertent Opening of a Steam Generator Relief or Safety Valve," states that the inadvertent opening of a SG relief or safety valve event in Mode 3 is bounded by the "Offsite Power Available" case analyzed in MODE 2 of the SLB event analyzed in Section 15.1.5. The SLB event analyzed in MODE 2 does not credit a turbine trip and therefore remains bounding for an event initiating in MODE 3. Therefore, the inadvertent opening of a SG relief or safety valve event is not impacted by the proposed CNS, MNS, and HNP TS change. The existing analyses presented in Section 15.1.4 of each station's UFSAR remain bounded by the Section 15.1.5 analyses.

3.1.4 Steam System Break The steam release caused by a rupture of a main steam line would result in an initial increase in steam flow, which decreases during the accident as the steam pressure falls. The energy removal from the RCS causes a reduction of coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in an insertion of positive reactivity.

Enclosure to RA-20-0207 Page 9 of 22 For CNS and MNS, the steam system piping rupture event is analyzed at "hot zero power (i.e.,

MODE 2) and "hot full power" (i.e., MODE 1) per UFSAR Section 15.1.5, "Steam System Piping Failure." The CNS and MNS UFSARs state that a potentially more limiting steam line break accident could occur for a steam line break outside containment when in MODE 3 with the low pressurizer signal for safety injection actuation blocked. In this scenario, feedwater would not automatically isolate, and return-to-power peak heat fluxes may increase to values significantly greater than the steam line break analysis initiating from no-load conditions. However, when safety injection is blocked, administrative controls on boron concentration are required for CNS and MNS to protect against such a return-to-power condition in MODE 3. The MODE 2 analysis does not credit a turbine trip and therefore remains bounding for the MODE 3 condition with administrative controls in place with the proposed change.

For HNP, four scenarios were considered in UFSAR Section 15.1.5, "Steam System Piping Failure." The reactor was assumed to be operating at either hot full power (i.e., MODE 1) or hot zero power (i.e., MODE 2), either with or without offsite power. The most limiting scenario was the hot zero power case with offsite power available. The HNP UFSAR states that during RCS cooldown and depressurization, administrative controls require for boration of the RCS to cold shutdown conditions prior to blocking the Safety Injection (SI) actuation signals. Thus, if a steamline rupture occurs with the SI actuation signals blocked, there would not be any return to criticality and the core would be protected. The MODE 2 analysis does not credit a turbine trip and therefore remains bounding for the MODE 3 condition with administrative controls in place with the proposed change.

Regardless of which mode the reactor is in at the start of the SLB transient (MODE 2 or MODE 3), initial steam flow is through the steam dumps to the condenser. Break flow through the break will dominate the small amount of steam flow through the steam dumps present at the start of the transient. There can be negligible steam flow through the turbine valves. Following MSIV closure in the MODE 2 analysis, the steam flow will not reach the turbine valves.

Therefore, a steam line break occurring in MODE 3 is bounded by the MODE 2 analysis and is not impacted by the proposed change. The existing hot zero power (MODE 2) with offsite power available analysis presented in UFSAR Section 15.1.5 for HNP remains bounding.

The events in Section 15.2 of the respective UFSARs address events that result in a decrease in heat removal by the secondary system. These events are discussed below in Section 3.1.5.

3.1.5 Decrease in Heat Removal by the Secondary System The consequences of a decrease in heat removal by the secondary system are discussed in the CNS, MNS and HNP UFSARs in Section 15.2, "Decrease in Heat Removal by the Secondary System." The category of the events includes:

(1) Loss of external electrical load (UFSAR Section 15.2.2),

(2) Turbine trip (UFSAR Section 15.2.3),

(3) Inadvertent closure of main steam isolation valves (UFSAR Section 15.2.4),

(4) Loss of condenser vacuum and other events resulting in turbine trip (UFSAR Section 15.2.5),

(5) Loss of offsite power (LOOP) (UFSAR Section 15.2.6, "Loss of Nonemergency AC

[Alternating Current] Power to the Station Auxiliaries"),

(6) Loss of normal feedwater (LONF) (UFSAR Section 15.2.7), and

Enclosure to RA-20-0207 Page 10 of 22 (7) Feedwater system pipe break (UFSAR Section 15.2.8).

These events are characterized by rapid reductions in heat removal capability of the SGs. The loss of heat removal capability results in a rapid rise in the SGs' secondary system pressure and temperature, and a subsequent increase in the RCS pressure and temperature. Reactor trip and actuation of primary and secondary safety valves mitigate the effects of the primary-to-secondary power mismatch during these events. The severity of these events is increased as the primary-to-secondary power mismatch is increased. The occurrence of this category of event at full power (i.e., MODE 1) produces a higher and more rapid power mismatch than at lower power operations or operations below MODE 2 because of a higher initial power and a higher decay heat level. Therefore, the worst cases are the events initiating from MODE 1 conditions.

For the above events (1) through (4), the turbine trip is the limiting event. Since this event is initiated by a turbine trip, the proposed deletion of the turbine trip function of the P-4 interlock in MODE 3 does not adversely affect these events. Thus, the analysis presented in Section 15.2.3, "Turbine Trip" of the CNS, MNS and HNP UFSARs remains bounding.

The LOOP and LONF events (5) and (6) are similar. Consequences of the initiating events are a reactor trip and Auxiliary Feedwater (AFW) actuation. Following a reactor trip, the decay heat (and reactor coolant pump heat for the LONF event) may initially exceed the heat removal capability of the secondary system. This will result in an increase in RCS pressure, temperature and pressurizer water level, and will continue until the AFW system re-establishes the secondary side heat sink and matches the decay heat load. The AFW system can remove stored and residual heat, thus preventing overpressurization of the RCS. Although the CNS, MNS and HNP UFSARs assume a turbine trip in MODE 1, delaying actuation of the turbine trip is a benefit to the results of the analyses, since the turbine would provide an additional heat removal path. The additional energy removed through the turbine would cause a less severe RCS overpressurization.

In MODE 3, the turbine is essentially already in a tripped state. The reactor is subcritical with lower decay heat at the start of the event. Both the primary and secondary system pressurization would clearly be bounded by these events initiated from a MODE 1 condition.

The existing MODE 1 analyses presented in Sections 15.2.6 and 15.2. 7 of the CNS, MNS and HNP UFSARs remain bounding.

Event (7), a feedwater system pipe break event, is defined as a break in a feedwater line large enough to prevent the addition of sufficient feedwater to the steam generators to maintain shell side fluid inventory. For breaks in certain locations, fluid from the SG may also be discharged through the break. A break could also preclude the subsequent addition of auxiliary feedwater to the affected SG. A reduced capacity of the secondary side heat sink results in RCS heatup due to the reduction in decay heat removal. The feedwater system line break event is limiting in MODE 1 because of the high decay heat level immediately following a reactor trip. Although the CNS, MNS and HNP MODE 1 analyses assume a turbine trip following a reactor trip, delaying the turbine trip actuation is a benefit since the turbine would provide an additional heat removal path.

In MODE 3, the turbine is essentially already in a tripped state. The reactor is subcritical with lower decay heat at the start of the event. Both the primary and secondary system pressurization would clearly be bounded by these events initiated from a MODE 1 condition.

Enclosure to RA-20-0207 Page 11 of 22 Furthermore, the MSIVs close in the analyses for all 3 plants thereby eliminating any steam from reaching the turbine following MSIV closure. The existing MODE 1 analyses presented in Section 15.2.8 of the CNS, MNS and HNP UFSARs remain bounding.

The events in Section 15.3 of the respective UFSARs address events that result in a decrease in reactor coolant flow. These events are discussed below in Section 3.1.6.

3.1.6 Decrease in Reactor Coolant Flow Section 15.3, "Decrease in Reactor Coolant System Flowrate" of the CNS, MNS and HNP UFSARs discusses the consequences of a decrease in RCS flow events. The applicable events are:

(1) Partial loss of forced reactor coolant flow (UFSAR Section 15.3.1 ),

(2) Complete loss of forced reactor coolant flow (UFSAR Section 15.3.2),

(3) Reactor coolant pump shaft seizure (locked rotor) (UFSAR Section 15.3.3), and (4) Reactor coolant pump shaft break (UFSAR Section 15.3.4).

For these events, the effect of a loss of coolant flow is a rapid increase in coolant temperature.

The heatup results in an increase in the RCS pressure and a decrease in the departure from nucleate boiling ratio (DNBR). The occurrence of events where there is a decrease in RCS flow at MODE 1 produces a higher and more rapid heatup than at lower power or operations in either MODE 2 or MODE 3. In MODE 3, the turbine is essentially already in a tripped state. The reactor is subcritical with lower decay heat at the start of the event. Both the primary and secondary system pressurization as well as core heat flux would clearly be bounded by these events initiated from a MODE 1 condition. Therefore, the events initiating from MODE 1 are limiting and result in a maximum peak RCS pressure and a minimum DNBR. Thus, the proposed change to remove the turbine trip function of the P-4 interlock in MODE 3 does not affect the decrease in reactor coolant flow events. The existing MODE 1 analyses presented in Sections 15.3.1 through 15.3.4 of the CNS, MNS and HNP UFSARs remain bounding.

Section 15.4 of the respective UFSARs address events caused by the addition of positive reactivity to the RCS. These events are addressed individually in Sections 3.1. 7 through 3.1.13 below.

3.1. 7 Uncontrolled Rod Cluster Control Assembly (RCCA) Bank Withdrawal from a Subcritical or Low Power Startup Condition A RCCA withdrawal accident is defined as an uncontrolled addition of reactivity to the reactor core caused by withdrawal of RCCAs resulting in a power excursion and is discussed in Section 15.4.1, "Uncontrolled Rod Cluster Control Assembly Bank Withdrawal From A Subcritical or Low Power Startup Condition" of the CNS, MNS and HNP UFSARs. The CNS, MNS and HNP UFSARs state that the reactor is conservatively assumed to be at hot zero power. Because none of the P-4 interlock functions are credited for the uncontrolled RCCA bank withdrawal from subcritical or lower power event, these functions are not required to mitigate the event.

Therefore, the proposed change to remove the turbine trip function of the P-4 interlock in MODE 3 does not affect this event. The existing MODE 2 analyses presented in Section 15.4.1 of the CNS, MNS and HNP UFSARs remain bounding.

Enclosure to RA-20-0207 Page 12 of 22 3.1.8 Uncontrolled RCCA Bank Withdrawal at Power As discussed in Section 15.4.2, "Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power," of the CNS, MNS and HNP UFSARs, the continuous uncontrolled RCCA bank withdrawal at power event is analyzed in MODE 1. Section 15.4.1 of the UFSARs addresses an uncontrolled RCCA bank withdrawal for MODES 2 and below and is discussed in the previous section. Thus, the proposed change to remove the turbine trip function of the P-4 interlock in MODE 3 does not affect this event. The analyses presented in Section 15.4.2 of the CNS, MNS and HNP UFSARs remain bounding.

3.1.9 Rod Cluster Control Assembly Misoperation The following events are considered RCCA misoperation accidents:

(1) One or more dropped RCCAs within the same group; (2) A dropped RCCA bank; (3) Statically misaligned RCCA; and (4) Withdrawal of a single RCCA.

These events result in core radial power distribution perturbations, which result in a decrease in the calculated DNBRs. Events (1) through (3) above are not limiting in MODE 3 because the core is not at power and a dropped or misaligned rod would not cause an adverse power distribution.

For event (4), withdrawal of a single RCCA, the analyses discussed in Section 15.4.3, "Rod Cluster Control Assembly Misoperation (System Malfunction or Operator Error)" of the CNS, MNS and HNP UFSARs do not credit the turbine trip P-4 interlock function to mitigate the event.

Thus, the proposed change to remove the turbine trip P-4 interlock function in MODE 3 does not affect the RCCA misoperation events. The existing analyses presented in Section 15.4.3 of the CNS, MNS and HNP UFSARs remain bounding.

3.1.10 Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature The HNP UFSAR 15.4.4 states that for MODE 1 and MODE 2 the event is not credible because the Plant Technical Specifications require that three reactor coolant pumps operate in MODES 1 and 2. In MODES 3 thru 6 the reactor is subcritical and there is no significant load on the plant.

The potential for a significant reactivity excursion following an inadvertent startup of an idle RCP is negligible. The consequences during these modes are bounded by those of Event 15.4.1 (Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from a Subcritical or Low Power Startup Condition).

For CNS and MNS the startup of an inactive reactor coolant pump at an incorrect temperature is analyzed in MODE 1 at 50% power. CNS/MNS analyses do not trip the reactor on any trip parameters. The P-4 interlock turbine trip function is not credited in this analysis.

The P-4 interlock functions, including the turbine trip on a reactor trip, are not credited in the event analyses for CNS, MNS and HNP. Thus, the proposed change to remove the turbine trip function of the P-4 interlock in MODE 3 does not affect the startup of an inactive RCP event.

The existing CNS, MNS, and HNP analyses presented in the respective UFSARs remain bounding with the proposed change.

Enclosure to RA-20-0207 Page 13 of 22 3.1.11 Chemical and Volume Control System (CVCS) Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant CVCS malfunctions result in a decrease in boron concentration in the reactor coolant. The resulting boron dilution event is analyzed in MODES 1 through 5. Since the analyses do not credit the P-4 interlock turbine trip on a reactor trip, the proposed change to remove the turbine trip function of the P-4 interlock in MODE 3 does not affect the results of the analyses.

Therefore, the existing analyses presented in Section 15.4.6 of the CNS, MNS and HNP UFSARs for this event remain bounding.

3.1.12 Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position Fuel and core loading errors (e.g., inadvertent loading of one or more fuel assemblies into improper positions) will lead to increased heat fluxes if the error results in placing fuel in core positions calling for fuel of lesser enrichment. Any error in enrichment, beyond the normal manufacturing tolerances, can cause power shapes which are more peaked than those calculated with the correct enrichments. The analyses presented in Section 15.4.7, "Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position" of the CNS, MNS and HNP UFSARs do not take credit for any of the P-4 interlock functions. Thus, the proposed change to remove the turbine trip function of the P-4 interlock in MODE 3 does not affect this event. The analyses presented in Section 15.4. 7 of the UFSARs remain bounding.

3.1.13 Spectrum of Rod Cluster Control Assembly Ejection Accidents The RCCA ejection event is the result of a mechanical failure of control rod mechanism pressure housing, resulting in the ejection of a RCCA and drive shaft. The event will result in a rapid positive reactivity insertion and system depressurization together, with an adverse core power distribution, possibly leading to localized fuel rod damage. The P-4 interlock functions, including the turbine trip on a reactor trip, are not credited in the RCCA ejection event analyses for CNS, MNS and HNP. Therefore, the proposed change to remove the turbine trip function of the P-4 interlock in MODE 3 does not affect the RCCA ejection event. The existing analyses presented in Section 15.4.8, "Spectrum of Rod Cluster Control Assembly Ejection Accidents" of the CNS, MNS and HNP UFSARs remain bounding.

Section 15.5 of the respective UFSARs address events caused by the addition of inventory to the RCS. These events are addressed individually in Sections 3.1.14 and 3.1.15 below.

3.1.14 Inadvertent Operation of the Emergency Core Cooling System During Power Operation An inadvertent Emergency Core Cooling System (ECCS) actuation at power results in an increase in RCS inventory, leading to the potential filling of the pressurizer. The event is terminated by the operator securing ECCS flow. The analyses in Section 15.5.1, "Inadvertent Operation of the Emergency Core Cooling System During Power Operation" of the CNS, MNS and HNP UFSARs evaluate core cooling capabilities and analyze pressurizer overfill. The CNS, MNS, and HNP UFSARs evaluated core cooling with a reactor trip assumed from a full power initial condition. The core cooling results are bounded by the inadvertent opening of a pressurizer safety or relief valve transient (See Section 3.1.16).

In MODE 3 the turbine is essentially in a tripped state with steam flows aligned to steam dumps.

Therefore, the proposed change to remove the turbine trip function of the P-4 interlock in MODE

Enclosure to RA-20-0207 Page 14 of 22 3 does not affect the inadvertent ECCS actuation at power event. The existing analyses presented in Section 15.5.1 of the CNS, MNS and HNP UFSARs remain bounding.

3.1.15 eves Malfunction that Increases Reactor Coolant Inventory Per Section 15.5.2, "Chemical and Volume Control System Malfunction that Increases Reactor Coolant Inventory" of the CNS, MNS and HNP UFSARs, an increase in reactor coolant inventory which results from the addition of cold, unborated water to the RCS is analyzed in UFSAR Section 15.4.6 (CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant Inventory). An increase in reactor coolant inventory which results from the injection of highly borated water into the RCS is analyzed in Section 15.5.1 (Inadvertent Operation of the ECCS During Power Operation). For a discussion of those UFSAR events with respect to the proposed change, refer to Sections 3.1.11 and 3.1.14 of this submittal.

The events in Section 15.6 of the respective UFSARs address events that result in a decrease in RCS inventory. These events are discussed individually below in Sections 3.1.16 through 3.1.19.

3.1.16 Inadvertent Opening of a Pressurizer Safety or Relief Valve An accidental depressurization of the RCS could occur as a result of an inadvertent opening of a pressurizer safety or relief valve. This event is analyzed in Section 15.6.1 of the CNS, MNS and HNP UFSARs, with the plant at MODE 1 conditions. During the transient, the RCS pressure decreases slowly. Although a turbine trip is modeled to occur on a reactor trip, the turbine trip is not used to mitigate the event. If this event were initiated from MODE 3, the turbine is essentially in a tripped state with the initial steam flow through the steam dump system. Steam dumps are not credited to mitigate the consequences of this accident. Thus, the turbine trip function of the P-4 interlock does not provide a mitigating effect for this event.

The proposed change to remove the turbine trip function of the P-4 interlock in MODE 3 does not affect the RCS depressurization event. The existing analyses presented in Section 15.6.1 of the CNS, MNS and HNP UFSARs remain bounding.

3.1.17 Break in Instrument Line or Other Lines from Reactor Coolant Pressure Boundary that Penetrate Containment For CNS and MNS, UFSAR Section 15.6.2 states that the most severe pipe rupture with regard to radioactivity release during normal plant operation occurs in the CVCS. This would be a complete severance, at rated power conditions (i.e., MODE 1 ), of the 3-inch letdown line just outside containment, between the outboard letdown isolation valve and the letdown heat exchanger. The occurrence of a complete severance of the letdown line would result in a loss of reactor coolant for CNS and MNS at the rate of approximately 140 gallons per minute and 150 gallons per minute respectively. Since the release rate is within the capability of the reactor makeup system, it would not result in Engineered Safety Features System actuation. Therefore, since the event is limiting in MODE 1 and the P-4 interlock function is not required for mitigation of the event, the proposed change does not affect the analysis of this event. The existing analyses presented in Section 15.6.2 of the CNS and MNS UFSARs remain bounding.

For HNP, UFSAR Section 15.6.2 states that the most severe pipe rupture with regard to radioactivity release during normal plant operation is a letdown line which penetrates

Enclosure to RA-20-0207 Page 15 of 22 containment. The occurrence of a complete severance of the letdown line would result in a loss of reactor coolant from HNP at a rate of approximately 200 gallons per minute. A turbine trip follows a reactor trip and results in an increase in secondary side pressure. The safety injection signal on low pressurizer pressure terminated the break flow by isolating the letdown line inside containment. In MODE 3, the turbine is essentially in a tripped condition with steam flow going through the steam dumps, so a turbine trip signal is not required to mitigate this event.

3.1.18 Steam Generator Tube Rupture The steam generator tube rupture (SGTR) event (CNS, MNS, HNP UFSAR 15.6.3) is analyzed from hot full power conditions to show that the consequences of radiological releases from the event are within the applicable dose limits. The radiological releases are calculated based on the mass releases from the SGTR transient analysis and site-specific meteorological parameters. Although a turbine trip is modeled to occur on a reactor trip, a safety injection signal will be generated either automatically or manually due to the continued depressurization of the RCS caused by the SGTR. This generates a separate turbine trip signal. In MODE 3, the turbine is essentially in a tripped condition, so a turbine trip signal is not required to mitigate this event. The mass releases during the mitigation of a SGTR event are driven by primary-to-secondary heat transfer which consists of stored energy in the primary system and core decay heat.

Both the initial stored energy and core decay heat levels are much less for a SGTR event initiating from a MODE 3 condition versus one initiating from a MODE 1 full power condition. The resultant steam mass releases to the environment would therefore continue to be bounded by the MODE 1 event. Therefore, since the event is limiting in MODE 1 and the P-4 interlock function is not required for mitigation of the event, the proposed change does not affect the analysis of this event.

3.1.19 LOCA Analysis - Large Break and Small Break Section 15.6.5 of the CNS, MNS and HNP UFSARs discusses the analyses of the small and large break loss-of-coolant (LOCA) accident. Since the analyses do not credit the turbine trip function of the P-4 interlock, the proposed change to remove the turbine trip function of the P-4 interlock in MODE 3 does not affect the LOCA accident. The existing analyses presented in Section 15.6.5 of the CNS, MNS and HNP UFSARs remain bounding.

The remainder of this discussion addresses the impact of removing the P-4 interlock in MODE 3 for events and analyses other than those in UFSAR Sections 15.1 through 15.6.

3.1.20 Anticipated Transients Without Scram An anticipated transient without scram (A TWS) is an anticipated operational occurrence followed by failure of the reactor trip portion of the reactor protection system. An A TWS is only postulated in MODE 1. Therefore, the proposed change to remove the turbine trip function of the P-4 interlock in MODE 3 does not affect the A TWS event. The existing analyses presented in Section 15.8 (ATWS) of the CNS, MNS and HNP UFSARs remain bounding.

Enclosure to RA-20-0207 Page 16 of 22 3.1.21a Long-Term (L T)/Short-Term (ST) LOCA Mass & Energy (M&E) (MNS/CNS)

The UFSAR Section 6.2.1.3.1.2 long-term M&E release analyses for postulated loss-of-coolant accidents assume that the turbine trips following the reactor trip, with a short delay included.

Modeling turbine trip minimizes heat removal from the reactor coolant system (RCS) primary side through the steam generators, other than from the steam generator safety valves, and forces the break to be the primary means of heat removal for the RCS.

The intent of the long term LOCA M&E releases analysis is to maximize the mass and energy release available to containment using limiting initial conditions and boundary conditions. None of these conditions, models or methodology used will become more limiting by implementing the proposed P-4 technical specification change. Therefore, the current design basis long-term LOCA mass and energy release analyses documented in the UFSAR would remain valid.

The UFSAR Chapter 6.2.1.3.1.1 short-term LOCA-related M&E release analysis results are used as input to the sub-compartment analyses, which are performed to ensure that the walls of the sub-compartments within the Reactor Building can maintain their structural integrity. There is a short pressure pulse in lower containment (less than 5 seconds) that accompanies a high energy line pipe rupture (LOCA) leading to significant pressure differentials across the walls between these sub-compartments. Due to this rapid timing, the P-4 interlock turbine trip function has no impact on the LOCA short-term mass and energy release analyses. Therefore, the current design basis short-term LOCA mass and energy releases analyses documented in the UFSAR would remain valid.

3.1.21b Long-Term (L T)/Short-Term (ST) LOCA Mass & Energy (M&E) (HNP)

For the UFSAR Chapter 6.2.1.3 LOCA M&E release analyses, initial system conditions are chosen to maximize the available energy in the system. Core power is assumed to be 102% of the Rated Thermal Power. Steam generator parameters are based on 100% full power. These conditions result in conservatively high M&E releases.

The intent of the long term LOCA M&E releases analysis is to maximize the mass and energy release available to containment using limiting initial conditions and boundary conditions. None of these conditions, models or methodology used will become more limiting by implementing the proposed P-4 technical specification change. Therefore, the current design basis analysis of record documented in the UFSAR would remain valid.

The short-term LOCA-related mass and energy releases are used as input to the sub-compartment analyses, which are performed to ensure that the walls of a sub-compartment can maintain their structural integrity. These analyses are performed to ensure that the walls in the immediate proximity of the break location can maintain their structural integrity during the short pressure pulse (generally less than 3 seconds) that accompanies a high energy line pipe rupture (LOCA) within the region (sub-compartment). Due to the short duration of the event, the effect from the P-4 interlock turbine trip function does not affect the LOCA short-term mass and energy release analyses. Therefore, the current design basis short-term LOCA mass and energy releases analysis documented in the UFSAR remains valid.

Enclosure to RA-20-0207 Page 17 of 22 3.1.22a Main Steam Line Break (MSLB) Inside Containment (IC) - M&E (CNS/MNS)

The UFSAR Chapter 6.2.1.4 Steam Line Break M&E release analysis for breaks inside containment assumes that the turbine trip is coincident with reactor trip, which results from a safety injection signal due to high containment pressure. All initial and boundary conditions (including the initial power level of 102% FP) are selected in order to maximize the enthalpy of the releases from the faulted generator.

These assumptions conservatively maximize the steam release out the break (rather than to the turbine). As such, the turbine trip on reactor trip is not credited in the analysis as a mitigating function. It is a penalty, not a benefit with respect to MSLB M&E releases. Additionally, MSIV closure occurs on high-high containment pressure closely following the reactor trip. After MSIV closure, a turbine trip would have no impact on containment response. Therefore, the P-4 interlock function has no impact on the MSLB M&E release analysis. The current design basis M&E release analyses documented in the UFSAR would remain valid.

3.1.22b Main Steam Line Break (MSLB) Inside Containment (IC) - M&E (HNP)

The UFSAR Chapter 6.2.1.4 MSLB M&E release is modeled to conservatively maximize the steam release out the break (rather than to the turbine). As such, the turbine trip on reactor trip is not credited in the analysis as a mitigating function. It is a penalty, not a benefit with respect to main steam line break mass and energy releases. Additionally, steam line isolation occurs closely following the reactor trip. After steam line isolation the turbine trip would have no impact on containment response. Therefore, the impact of the P-4 interlock functions has no impact on the MSLB M&E release analysis. The current design basis M&E release analyses documented in the UFSAR would remain valid.

3.2 Steam Dump Control (Catawba ONLY)

One of the functions of the P-4 interlock that CNS currently has in TS is to transfer the steam dump from the load rejection controller to the unit trip controller. The MNS and HNP steam controllers perform the same function, using different signals, without a notation in TS.

CNS UFSAR Section 7.7, "Control Systems Not Required for Safety," identifies Steam Dump Control as a plant control system not required for safety. The general design objectives of the plant control systems are:

1. To establish and maintain power equilibrium between primary and secondary system during steady state unit operation;

2. To constrain operational transients so as to preclude unit trip and re-establish steady state unit operation;

3. To provide the reactor operator monitoring instrumentation that indicates all required input and output control parameters of the systems and provides the operator the capability of assuming manual control of the system.

Enclosure to RA-20-0207 Page 18 of 22 More specifically with respect to Steam Dump Control, the system:

Permits the unit to accept a sudden loss of load without incurring a reactor trip. Steam is dumped to the condenser and/or the atmosphere, as necessary, to accommodate excess power generation in the reactor during turbine load reduction transients.

Ensures that stored energy and residual heat are removed following a reactor trip to bring the plant to equilibrium no-load conditions, without actuation of the steam generator safety valves.

Maintains the plant at no-load conditions and permits a manually controlled cooldown of the plant.

The CNS TS 3.3.2 Bases states, in part, that the ESFAS instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). 10 CFR 50.36(c)(2)(ii) requires that (c) technical specifications will include items in the following categories: (2) limiting conditions for operation: (ii) a technical specification limiting condition for operation must be established for each item meeting one or more of the following criterion: Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The instrumentation utilized to initiate transfer to the unit trip steam dump controller does not serve a primary protective function to warrant inclusion in the TS. The instrumentation does not serve to ensure that the plant is operated within the bounds of initial conditions assumed in design basis accident and transient analyses. Likewise, the transfer to the unit trip steam dump controller instrumentation does not serve as part of the primary success path of a safety analysis used to demonstrate that the consequences of these events are within the appropriate acceptance criteria.

The specific instrumentation requirements related to transfer to the unit trip steam dump controller are not required to be in the CNS TS based on the criteria in 10 CFR 50.36 and are not required to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety. Therefore, this function of the P-4 interlock is excluded from the proposed footnote (h) in the "APPLICABLE MODES AND OTHER SPECIFIED CONDITIONS" portion of CNS TS Table 3.3.2-1, Function 8.a.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The Commission's regulatory requirements related to the content of TS are set forth in Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36, "Technical specifications." This regulation requires that the TS include items in five categories. These categories include (1) safety limits, limiting safety system settings, and limiting control setting, (2) limiting conditions for operation (LCOs), (3) surveillance requirements, (4) design features, and (5) administrative controls.

The requirements in 10 CFR 50.36(c)(2)(ii) set forth four criteria to be used in determining whether a LCO is required to be included in TS. These criteria are:

Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Enclosure to RA-20-0207 Page 19 of 22 Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

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

Criterion 4. A structure, system, or component which operating experience or probablistic risk assessment has shown to be significant to public health and safety.

General Design Criterion (GDC) 13, "lnstrumenation and control," of Appendix A to 10 CFR Part 50, requires that appropriate controls be provided to monitor variables and systems over their anticipated ranges for normal operation, for AOOs, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems within prescribed operating ranges.

GDC 20, "Protection system functions," of Appendix A to 10 CFR Part 50, requires that the protection system be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

GDC 21, "Protection system reliability and testability," of Appendix A to 10 CFR Part 50, requires that the protection system be designed and tested for high functional reliability.

The proposed change to remove the turbine trip function of the P-4 interlock in MODE 3 out of the CNS, MNS and HNP TS is consistent with the CNS, MNS and HNP designs and analyses and will continue to ensure proper actuation to satisfy the anticipatory trip function. Therefore, the proposed change for CNS, MNS and HNP does not affect compliance with the above regulations and will ensure the lowest functional capabilities or performance levels of equipment required for safe operation are met.

4.2 Precedent The proposed change is consistent with NRG-approved license amendment issued to Wolf Creek Nuclear Operating Corporation on March 30, 2011 (Amendment No. 194) for Wolf Creek Generating Station (ADAMS Accession No. ML110550846). The approved TS change regarding removal of the turbine trip function of the P-4 interlock in MODE 3 is identical to the change proposed in this request for CNS, MNS and HNP. The approved TS change regarding removal of the P-4 function related to the steam dump function in MODES 1, 2 and 3 is identical to the change proposed in this request for CNS.

4.3 No Significant Hazards Consideration Determination Analysis Duke Energy is requesting amendments to the Technical Specifications (TS) for Catawba Nuclear Station (CNS), Units 1 and 2; McGuire Nuclear Station (MNS), Units 1 and 2; and Shearon Harris Nuclear Power Plant (HNP), Unit 1.

Enclosure to RA-20-0207 Page 20 of 22 For CNS and MNS, the proposed change would revise Table 3.3.2-1, Function 8.a. (ESFAS Interlocks, Reactor Trip, P-4) of TS 3.3.2, "Engineered Safety Feature Actuation System (ESFAS) Instrumentation." For HNP, the proposed change would revise Table 3.3-3, Functional Unit 1 0.c. (Engineered Safety Features Actuation System Interlocks, Reactor Trip, P-4) and Table 4.3-2, Channel Functional Unit 1 0.c. (Engineered Safety Features Actuation System Interlocks, Reactor Trip, P-4) of TS 3/4.3.2, "Engineered Safety Features Actuation System Instrumentation." Duke Energy is proposing to add a footnote to the applicable modes portion of the P-4 interlock Function to identify the enabled functions (i.e., sub-functions of the interlock required to meet the Limiting Condition for Operation) and the applicable MODES for each enabled function.

For CNS, MNS and HNP, the proposed change, in effect, removes the turbine trip function of the P-4 interlock in MODE 3 from TS. For CNS only, the proposed change also removes the P-4 interlock function of transferring the steam dump from the load rejection controller to the unit trip controller out of TS.

Duke Energy has evaluated whether or not a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

Overall protection system performance will remain within the bounds of the previously performed accident analyses for CNS, MNS and HNP. Bypassing the turbine trip on reactor trip function in MODE 3 will not impact any accidents previously evaluated in the Updated Final Safety Analysis Reports (UFSARs) for CNS, MNS and HNP.

Regarding the proposed change for CNS only, the instrumentation utilized to initiate transfer to the unit trip steam dump controller does not serve a primary protective function to warrant inclusion in the TS. The instrumentation does not serve to ensure that the plant is operated within the bounds of initial conditions assumed in design basis accident and transient analyses. Likewise, the transfer to the unit trip steam dump controller instrumentation does not serve as part of the primary success path of a safety analysis used to demonstrate that the consequences of these events are within the appropriate acceptance criteria.

The ESFAS at CNS, MNS and HNP will continue to function in a manner consistent with the accident analyses assumptions and the respective plant design basis. As such, there will be no degradation in the performance of, nor an increase in the number of challenges to, equipment assumed to function during an accident situation. The proposed change to the CNS, MNS and HNP TS does not affect the probability of any event initiators. There will be no change to normal plant operating parameters or accident mitigation capabilities.

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

Enclosure to RA-20-0207 Page 21 of 22

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

Response: No.

There are no changes in the method by which any safety related plant system at CNS, MNS and HNP performs its safety function and the normal manner of plant operation at CNS, MNS and HNP is unaffected, other than the proposed allowance to defeat the turbine trip on reactor trip function of the P-4 interlock in MODE 3.

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. There will be no adverse effects or challenges imposed on any safety related systems as a result of this proposed change.

Therefore, the possibility of a new or different type of accident is not created.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

There will be no effect on the way safety limits or limiting safety system settings are determined, nor will there be any effect on those plant systems necessary to assure the accomplishment of protection functions. There will be no impact on departure from nucleate boiling ration limits, heat flux hot channel factor limits, nuclear enthalpy rise hot channel factor limits, peak centerline temperature limits, peak local power density or any other margin of safety.

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

Based on the above, Duke Energy concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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

Enclosure to RA-20-0207 Page 22 of 22 5.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

to RA-20-0207 Proposed Technical Specification Changes (Mark-up)

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 6 of 6)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

7. Automatic Switchover to Containment Sump

a.

Automatic 1,2,3,4 2 trains C

SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.4 and Actuation SR 3.3.2.6 Relays

b.

Refueling Water 1,2,3,4 4

N SR 3.3.2.1

91.9 inches 95 inches Storage Tank SR 3.3.2. 7(a)(b)

(RWST) Level -

SR 3.3.2.9(a)(b)

Low SR 3.3.2.10 Coincident with Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

Safety Injection

8.

ESFAS Interlocks

a.

Reactor 1,2,3l!!l 1 per train, F

SR 3.3.2.8 NA NA Trip, P-4 2 trains

b.

Pressurizer 1,2,3 3

0 SR 3.3.2.5

1944 and 1955 psig Pressure, P-11 SR 3.3.2.9

s; 1966 psig C.

Tavg - Low Low, 1,2,3 1 per loop 0

SR 3.3.2.5

550°F 553°F P-12 SR 3.3.2.9

9. Containment Pressure Control System

a.

Start Permissive 1,2,3,4 4 per train p

SR 3.3.2.1

s; 1.0 psid 0.9 psid SR 3.3.2.7 SR 3.3.2.9

b.

Termination 1,2,3,4 4 per train p

SR 3.3.2.1

0.25 psid 0.35 psid SR 3.3.2.7 SR 3.3.2.9

10. Nuclear Service 1,2,3,4 3 per pit Q,R SR 3.3.2.1

El. 555.4 ft El. 557.5 ft Water Suction SR 3.3.2.9 Transfer - Low Pit SR 3.3.2.11 Level SR 3.3.2.12 (a) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(b) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (field setting) to confirm channel performance. The methodologies used to determine the as-found and the as-left tolerances are specified in the UFSAR.

fh} The functions of the Reactor Trip P:4 interlock required to meet the LCO are*

Trip the main turbine MODES 1 and 2 Isolate MEW with coincident low Lw MODES 1 2 and 3 Prevent reactuation of SI after a manual reset of SI MODES 1 2 and 3 Prevent opening of MEIYs if closed on SI or SG Water Level - High High - MODES 1 2 and 3 Catawba Units 1 and 2 3.3.2-18 Amendment Nos. 2771273

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 6 of 6)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

8.

ESFAS Interlocks

a.

Reactor Trip, 1,2,3ill 1 per train, F

SR 3.3.2.7 NA NA P-4 2 trains

b.

Pressurizer 1,2,3 3

Q SR 3.3.2.5

~ 1965 psig 1955 psig Pressure, P-11 SR 3.3.2.8 C.

Tavg - Low Low, 1,2,3 1 per loop Q

SR 3.3.2.5

~ 551 °F 553°F P-12 SR 3.3.2.8

9.

Containment 1,2,3,4 4 per train, R

SR 3.3.2.1 Refer to Note Refer to Note Pressure Control 2 trains SR 3.3.2.3 1 on Page 1 on page System SR 3.3.2.8 3.3.2-14 3.3.2-14 lf\\ The functions of the Reactor Trip P-4 interiock required to meet the L co are*

Trip the main turbine - MODES 1 and 2 Isolate MEW with coincident low Lw - MODES 1 2 and 3 Prevent reactuation of SI after a manual reset of SI - MODES 1 2 and 3 Prevent opening of MEIYs if closed on SI or SG Water L evel - High High - MODES 1 2 and 3 McGuire Units 1 and 2 3.3.2-15 Amendment Nos. 272/252

TABLE 3.3-3 {Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. OF CHANNELS FUNCTIONAL UNIT CHANNELS TO TRIP CHANNELS OPERABLE APPLICABLE MODES ACTION

8.

Containment Spray Switch-over to Containment Sump (Continued)

b. RWST--Low Low Coincident With Containment Spray

9.

Loss-of-Offsite Power

a. 6.9 kV Emergency Bus--Undervoltage Primary

b. 6.9 kV Emergency Bus--Undervoltage Secondary

10.

Engineered Safety Features Actuation System Interlocks

a. Pressurizer Pressure, P-11 Not P-11

b. Low-Low Tavg, P-12

c. Reactor Trip, P-4

d. Steam Generator Water Level, P-14 SHEARON HARRIS - UNIT 1 See Item 7.b. above for all RWST--Low Low initiating functions and requirements.

See Item 2 above for all Containment Spray initiating functions and requirements.

3/bus 2/bus 2/bus 1, 2, 3,4 15a 3/bus 2/bus 2/bus 1, 2, 3,4 15a 3

2 2

1, 2, 3 20 3

2 2

1, 2, 3 20 3

2 2

1, 2, 3 20 2

2 2

1. 2. 31## 1 22 See Item 5.b. above for all P-14 initiating functions and requirements.

3/4 3-25 Amendment No. 4-79 I

TABLE 3.3-3 (Continued)

TABLE NOTATIONS

Trip function may be blocked in this MODE below the P-11 (Pressurizer Pressure Interlock)

Setpoint.

- During CORE AL TERA TIONS or movement of irradiated fuel in containment, refer to Specification 3.9.9.

- - Trip function automatically blocked above P-11 and may be blocked below P-11 when Safety Injection on low steam line pressure is not blocked.

INSERT 3/4 3-26 ACTION STATEMENTS ACTION 14 -

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE.

ACTION 15 -

With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until performance of the next required CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 15a -With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . With less than the minimum channels OPERABLE, operation may proceed provided the minimum number of channels is restored within one hour, otherwise declare the affected diesel generator inoperable. When performing surveillance testing of either primary or secondary undervoltage relays, the redundant emergency bus and associated primary and secondary relays shall be OPERABLE.

ACTION 16 -

With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1.

ACTION 17 -

With less than the Minimum Channels OPERABLE requirement, operation may continue provided the Containment Purge Makeup and Exhaust Isolation valves are maintained closed while in MODES 1, 2, 3 and 4 (refer to Specification 3.6.1. 7). For MODE 6, refer to Specification 3.9.4.

ACTION 18 -

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SHEARON HARRIS - UNIT 1 3/4 3-26 Amendment No. 4+9

CHANNEL FUNCTIONAL UNIT

10. Engineered Safety Features Actuation System Interlocks (Continued)

c. Reactor Trip, P-4 TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL CHECK CALIBRATION N.A.

N.A.

ANALOG CHANNEL OPERATIONAL TEST N.A.

TRIP ACTUATING DEVICE OPERATIONAL TEST SFCP MASTER SLAVE ACTUATION RELAY RELAY LOGIC TEST TEST TEST N.A.

N.A.

MODES FOR WHICH SURVEILLANCE IS REQUIRED 1, 2, 31## 1

d. Steam Generator Water See Item 5.b., above for P-14 Surveillance Requirements.

Level, P-14 SHEARON HARRIS - UNIT 1 3/4 3-48 Amendment No. 4a4

(1)

(2)

(3)

(4)

TABLE 4.3-2 (Continued)

TABLE NOTATION Each train shall be tested at the frequency specified in the Surveillance Frequency Control Program.

The Surveillance Requirements of Specification 4.9.9 apply during CORE ALTERATIONS or movement of irradiated fuel in containment.

Except for relays K601, K602, K603, K608, K610, K615, K616, K617, K622, K636, K739, K7 40 and K7 41 which shall be tested at the frequency specified in the Surveillance Frequency Control Program and during each COLD SHUTDOWN exceeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months unless they have been tested within the previous 92 days.

The Steam Line Isolation-Safety Injection (Block-Reset) switches enable the Negative Steam Line Pressure Rate--High signal (item 4.e) when used below the P-11 setpoint.

Verify proper operation of these switches each time they are used.

Setpoint verification not required.

During CORE AL TERA TIONS or movement of irradiated fuel in containment.

Trip Function automatically blocked above P-11 and may be blocked below P-11 when safety injection or low steamline pressure is not blocked.

INSERT 3/4 3-49 SHEARON HARRIS - UNIT 1 3/4 3-49 Amendment No. 4a4

INSERT 3/4 3-26

1. The functions of the Reactor Trip, P-4 interlock required to meet the LCO are:

Trip the main turbine - MODES 1 and 2 Isolate Main Feedwater with coincident low Tavg - MODES 1, 2, and 3 Prevent reactuation of Safety Injection after a manual reset of Safety Injection - MODES 1, 2, and 3 Prevent opening of Main Feedwater valves if closed on Safety Injection or Steam Generator Water Level - High High - MODES 1, 2, and 3 INSERT 3/4 3-49 The functions of the Reactor Trip, P-4 interlock required to meet the LCO are:

Trip the main turbine - MODES 1 and 2 Isolate Main Feedwater with coincident low Tavg - MODES 1, 2, and 3 Prevent reactuation of Safety Injection after a manual reset of Safety Injection - MODES 1, 2, and 3 Prevent opening of Main Feedwater valves if closed on Safety Injection or Steam Generator Water Level - High High - MODES 1, 2, and 3 to RA-20-0207 Proposed Technical Specification Bases Changes (Mark-up)

(For Information Only)

!No changes this page. I BASES ESFAS Instrumentation B 3.3.2 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

8.

Catawba Units 1 and 2 provide no control functions. Since an inadvertent switchover to the containment sump could have a significant safety impact, this instrumentation is placed in a bypass condition for testing. Therefore, four channels are supplied such that, during testing, the remaining three channels could perform the intended function, and no single failure could result in either a failure to accomplish the intended function, or in an inadvertent switchover to the containment sump.

Automatic switchover occurs only if the RWST low level signal is coincident with SI. This prevents accidental switchover during normal operation. Accidental switchover could damage ECCS pumps if they are attempting to take suction from an empty sump. The automatic switchover Function requirements for the SI Functions are the same as the requirements for their SI function. Therefore, the requirements are not repeated in Table 3.3.2-1. Instead, Function 1, SI, is referenced for all initiating Functions and requirements.

These Functions must be OPERABLE in MODES 1, 2, 3, and 4 when there is a potential for a LOCA to occur, to ensure a continued supply of water for the ECCS pumps.

These Functions are not required to be OPERABLE in MODES 5 and 6 because there is adequate time for the operator to evaluate unit conditions and respond by manually starting systems, pumps, and other equipment to mitigate the consequences of an abnormal condition or accident. System pressure and temperature are very low and many ESF components are administratively locked out or otherwise prevented from actuating to prevent inadvertent overpressurization of unit systems.

Engineered Safety Feature Actuation System Interlocks To allow some flexibility in unit operations, several interlocks are included as part of the ESFAS. These interlocks permit the operator to block some signals, automatically enable other signals, prevent some actions from occurring, and cause other actions to occur. The interlock Functions back up manual actions to ensure bypassable functions are in operation under the conditions assumed in the safety analyses.

B 3.3.2-25 Revision No. 12

BASES ESFAS Instrumentation B 3.3.2 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

a.

Engineered Safety Feature Actuation System Interlocks-Reactor Trip, P-4 The P-4 interlock is enabled when a reactor trip breaker (RTB) and its associated bypass breaker is open. Operators are able to reset SI 60 seconds after initiation. If a P-4 is present when SI is reset, subsequent automatic SI initiations will be blocked until the RTBs have been manually closed.

....-----------. This Function allows operators to take manual control of SI

-IN_ S_E_R_T_B---r-3_.3_._2_-2_6_

systems after the initial phase of injection is complete while Catawba Units 1 and 2 avoiding multiple SI initiations. The funGtions of the P 4 interlock are:

Trip the main turbine; Isolate MFVV with coincident 101.v TdV§-i-Pre11ent reaGtuation of SI after a manual reset of SI; Transfer the steam dump from the load rejeGtion controller to the unit trip controller; and Prevent opening of the MFW isolation valves if they 1.vere closed on SI or SG V\\later Le1,el l=ligh l=ligh.

Each of the above FunGtions functions is interlocked with P-4 to avert or reduce the continued cooldown of the RCS following a reactor trip. AR-A reactor trip from MODE 1 or 2 could result in an excessive cooldown of the RCS following a reactor tripthat could cause an insertion of positive reactivity with a subsequent increase in generated power. To avoid such a situation, the noted FunGtions functions have been interlocked with P-4 as part of the design of the unit control and protection system.

None of the noted Functions serves a mitigation function in the unit licensing basis safety analyses. Only the turbine trip FunGtion is mcplicitly assumed since it is an immediate consequence of the reaGtor trip FunGtion. Neither turbine trip, nor any of the other fuur FunGtions associated 11,ith the reactor trip signal, is required to show that the unit licensing basis safety analysis acceptance criteria are not e><ceeded.

The RTB position switches that provide input to the P-4 interlock only function to energize or de-energize or open or B 3.3.2-26 Revision No. 4-2

BASES ESFAS Instrumentation B 3.3.2 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

b.

C.

Catawba Units 1 and 2 close contacts. Therefore, this Function has no adjustable trip setpoint with which to associate a Trip Setpoint and Allowable Value.

This Function must be OPERABLE in MODES 1, 2, and 3 when the reactor may be critical or approaching criticality.

This Function does not have to be OPERABLE in MODE 4, 5, or 6 because the main turbine, the MFW System, and the Steam Dump System are not in operation.

Engineered Safety Feature Actuation System Interlocks-Pressurizer Pressure. P-11 The P-11 interlock permits a normal unit cooldown and depressurization without actuation of SI or main steam line isolation. With two-out-of-three pressurizer pressure channels (discussed previously) less than the P-11 setpoint, the operator can manually block the Pressurizer Pressure-Low SI signal and the Steam Line Pressure-Low steam line isolation signal (previously discussed). When the Steam Line Pressure-Low steam line isolation signal is manually blocked, a main steam isolation signal on Steam Line Pressure-Negative Rate-High is enabled. This provides protection for an SLB by closure of the MSIVs. With two-out-of-three pressurizer pressure channels above the P-11 setpoint, the Pressurizer Pressure-Low SI signal and the Steam Line Pressure-Low steam line isolation signal are automatically enabled. The operator can also enable these trips by use of the respective manual reset buttons. When the Steam Line Pressure-Low steam line isolation signal is enabled, the main steam isolation on Steam Line Pressure-Negative Rate-High is disabled.

This Function must be OPERABLE in MODES 1, 2, and 3 to allow an orderly cooldown and depressurization of the unit without the actuation of SI or main steam isolation. This Function does not have to be OPERABLE in MODE 4, 5, or 6 because system pressure must already be below the P-11 setpoint for the requirements of the heatup and cooldown curves to be met.

Engineered Safety Feature Actuation System lnterlocks-Tav9-Low Low, P-12 On increasing reactor coolant temperature, the P-12 interlock B 3.3.2-27 Revision No. 2

INSERT B 3.3.2-26 The functions of the P-4 interlock are:

Function Reguired MODE Trip the main turbine 1, 2 Isolate MFW with coincident low Tavg 1, 2, 3 Prevent reactuation of SI 1, 2, 3 after a manual reset of SI Transfer the steam dump from the load rejection None controller to the unit trip controller; and Prevent opening of the MFW isolation valves if 1, 2, 3 they were closed on SI or SG Water Level - High High

BASES ESFAS Instrumentation B 3.3.2 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

8.

there is adequate time for the operator to evaluate unit conditions and respond by manually starting systems, pumps, and other equipment to mitigate the consequences of an abnormal condition or accident. System pressure and temperature are very low and many ESF components are administratively locked out or otherwise prevented from actuating to prevent inadvertent overpressurization of unit systems.

Engineered Safety Feature Actuation System Interlocks To allow some flexibility in unit operations, several interlocks are included as part of the ESFAS. These interlocks permit the operator to block some signals, automatically enable other signals, prevent some actions from occurring, and cause other actions to occur. The interlock Functions back up manual actions to ensure bypassable functions are in operation under the conditions assumed in the safety analyses.

a.

Engineered Safety Feature Actuation System Interlocks-Reactor Trip. P-4 The P-4 interlock is enabled when a reactor trip breaker (RTB) and its associated bypass breaker is open. Operators are able to reset SI 60 seconds after initiation. If a P-4 is present when SI is reset, subsequent automatic SI initiation will be blocked until the RTBs

.---------..... have been manually closed. This Function allows operators to

~IN_S_ E_R_T_B~ 3_.3_._2_-2_3_

take manual control of SI systems after the initial phase of injection McGuire Unit 1 and 2 is complete while avoiding multiple SI initiations. The functions of the P 4 interlock are:

Trip the main turbine; Isolate MFW 1.¥ith coincident low T..vi,-

Pre11ent reaciuation of SI after a manual reset of SI; and

Prevent opening of the MFW isolation valves if they were closed on SI or SG Water Level High High.

B 3.3.2-23 Revision No. ~

BASES ESFAS Instrumentation B 3.3.2 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

b.

McGuire Unit 1 and 2 Each of the above Functions functions is interlocked with P-4 to avert or reduce the continued cooldown of the RCS following a reactor trip. AR-A reactor trip from MODE 1 or 2 could result in an excessive cooldown of the RCS following a reactor trip that could cause an insertion of positive reactivity with a subsequent increase in generated power. To avoid such a situation, the noted Functions functions have been interlocked with P-4 as part of the design of the unit control and protection system.

None of the noted Functions serves a mitigation function in the unit licensing basis safety analyses. Only the turbine trip Function is e>Eplicitly assumed since it is an immediate consequence of the reactor trip Function. Neither turbine trip, nor any of the other three Functions associated with the reactor trip signal, is required to show that the unit licensing basis safety analysis acceptance criteria are not e>Eceeded.

The RTB position switches that provide input to the P-4 interlock only function to energize or de-energize or open or close contacts.

Therefore, this Function has no adjustable trip setpoint with which to associate a Trip Setpoint and Allowable Value.

This Function must be OPERABLE in MODES 1, 2, and a when the reactor may be critical or approaching criticality. This Function does not have to be OPERABLE in MODE 4, 5, or 6 because the main turbine, the MFW System are not in operation.

Engineered Safety Feature Actuation System Interlocks-Pressurizer Pressure, P-11 The P-11 interlock permits a normal unit cooldown and depressurization without actuation of SI or main steam line isolation. With two-out-of-three pressurizer pressure channels (discussed previously) less than the P-11 setpoint, the operator can manually block the Pressurizer Pressure-Low SI signal and the Steam Line Pressure-Low steam line isolation signal (previously discussed).

B 3.3.2-24 Revision No. ~

INSERT B 3.3.2-23 The functions of the P-4 interlock are:

Function Reguired MODE Trip the main turbine 1, 2 Isolate MFW with coincident low Tavg 1, 2, 3 Prevent reactuation of SI 1, 2, 3 after a manual reset of SI Prevent opening of the MFW isolation valves if 1, 2, 3 they were closed on SI or SG Water Level - High High

3/4.3 INSTRUMENTATION BASES Table 3.3-4 includes values for 6.9 kV Emergency Bus Undervoltage - Secondary (degraded grid) trip setpoints and allowable values. The secondary undervoltage relays are connected to two distinct time delay relays. Upon expiration of the first time delay, which is long enough to accommodate the starting of the motor which has the longest starting time, an alarm is actuated at the main control board to alert the operator of this condition and to permit operator actions to restore the system voltage. Automatic tripping actions as described for the primary protection are initiated if a safety actuation signal is present after the expiration of the time delay.

In the event of a coincident large break loss of coolant accident (LBLOCA) and voltage dropping to actuate the short-term DVR function (bus voltage drops into the range between the DVR dropout voltage setting and the loss of offsite power voltage setpoint), a safety injection actuation signal is generated, emergency loads begin to sequence onto the emergency buses (still powered from the normal offsite supply), and the emergency diesel generator starts but does not load. If the degraded voltage condition continues to exist until the short-term DVR time delay setting is reached, the emergency loads are then separated from offsite power, loads on emergency buses are shed, the emergency diesel generator output breaker is shut, and the emergency loads are sequenced back onto the emergency buses. The LBLOCA analysis timeline for the safety functions provided by the equipment in this scenario is used to establish the analytical limit for the maximum short-term DVR time delay. This meets the intent of Branch Technical Position PSB-1 regarding maximum time delays consistent with design basis accident analysis.

If degraded voltage conditions exist without a simultaneous accident (normal operating conditions), a longer time delay (Device 2-2) is allowed before the automatic tripping actions are initiated. This second time delay is based on the maximum time for which the most sensitive load can perform its safety function without impairment at the degraded voltage.

Calculations to determine time delay allowable values and trip setpoints to protect time delay analytical limits were performed consistent with the methodology of Technical Specification Task Force Traveler 493, Clarify Application of Setpoint Methodology for LSSS Functions. Although the DVR function is not a limiting safety system setting function, the methodology is a conservative approach for determination of these parameters.

The Engineered Safety Features Actuation System interlocks perform the following functions:

P-4 Reaotor tripped /\\Gtuates Turbine trip, sloses rnain feedwater Yalves on Tavo belo\\*.i INSERT B 3/4 3-3 Setpoint, J:>FO¥ents the oponinJ of the R=1ain Jeedwater val>,<<es v.<hish were closed by a Safety lnjeotion or t=ligh Stearn Generator Water LeYel signal, allows Safety lnjeotion blosk so that sornponents san be reset or tripped.

Reactor not tripped - prevents manual block of Safety Injection.

P-11 On increasing pressurizer pressure, P-11 automatically reinstates Safety Injection actuation on low pressurizer pressure and low steam-line pressure, sends an open signal to the accumulator discharge valves and automatically blocks steam-line isolation on a high rate of decrease in steam-line pressure. On decreasing pressurizer pressure, P-11 allows the manual block of Safety Injection on low pressurizer pressure and low steam-line pressure and allows steam-line isolation, on a high rate of decrease in steam-line pressure, to become active upon manual block of Safety Injection from low steam-line pressure.

P-12 P-12 has no ESF or reactor trip functions. On decreasing reactor coolant loop temperature, P-12 automatically removes the arming signal from the Steam Dump System.

SHEARON HARRIS - UNIT 1 B 3/4 3-3

/\\rnendrnent No. 14 a

INSERT B 3/4 3-3 P-4 Reactor tripped -Actuates turbine trip (MODES 1 and 2), closes main feedwater valves on Tavg below setpoint (MODES 1, 2, and 3), prevents the opening of the main feedwater valves which were closed by a Safety Injection or High Steam Generator Water Level signal (MODES 1, 2, and 3), allows Safety Injection block so that components can be reset or tripped (MODES 1, 2, and 3).

v t e Letter Sequence Request
EPID:L-2020-LLA-0262, TSTF-493 (Approved, Closed)
Initiation Request Acceptance Administration Questions Answers Results Approval Other: ML22012A419
MONTHYEAR2021-01-06 [Table View]

RA-20-0207, Duke Energy - Application to Revise the Technical Specification for Engineered Safety Feature Actuation System (ESFAS) Instrumentation

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SUBJECT:

Enclosure:

Subject:

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

SUMMARY

3.0 TECHNICAL EVALUATION

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Person / Time
ML20338A264
Site:	Mcguire, Catawba, Harris, McGuire (NPF-009, NPF-017, NPF-035, NPF-052, NPF-063)
Issue date:	12/03/2020
From:	Snider S Duke Energy Carolinas, Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-20-0207
Download: ML20338A264 (43)
v • d • e

RA-20-0207, Duke Energy - Application to Revise the Technical Specification for Engineered Safety Feature Actuation System (ESFAS) Instrumentation

Text

SUBJECT:

Enclosure:

Subject:

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

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