Text

NRR E-mail Capture - Request for Additional Information - St. Lucie RICT LAR - MF5372/MF5373
	ML17277A369
Person / Time
Site:	Saint Lucie
Issue date:	10/04/2017
From:	Perry Buckberg; Plant Licensing Branch II
To:	Frehafer K; Florida Power & Light Co
References
	MF5372, MF5373
	Download: ML17277A369 (26)
	v • d • e

1 NRR-PMDAPEm Resource From:

Buckberg, Perry Sent:

Wednesday, October 04, 2017 11:36 AM To:

Frehafer, Ken Cc:

Kilby, Gary; Snyder, Mike; Mack, Jarrett (Jarrett.Mack@fpl.com)

Subject:

Request for Additional Information - St. Lucie RICT LAR - MF5372/MF5373 Attachments:

St Lucie EEOB RAIs RICT August 2017.pdf; St Lucie STSB RAIs RICT August 2017.pdf; St Lucie APLA RICT RAIs October 2017.pdf; DRAFT St Lucie STSB RAIs 505-RICT July 2017

.docx

Ken, By letter dated December 5, 2014, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14353A016), as supplemented by letters dated July 8, 2016, July 22, 2016 and February 25, 2017, Florida Power &

Light (FPL), submitted a license amendment request (LAR) for St. Lucie Units 1 and 2. The LAR proposes changes to the Technical Specifications (TSs) for Units 1 and 2 that would permit the use of a Risk Informed Completion Time (RICT) for several TSs. The NRC staff finds that additional information is needed to complete our review of the proposed amendment.

The NRC staffs Final Requests for Additional Information (RAIs) related to Electrical (EEOB), Tech Specs (STSB), and Probabilistic Risk Assessment (APLA) are attached. Draft copies of these RAIs were provided to you in July and August of 2017, and based on a clarification call held on August 9, 2017, the STSB RAIs were modified for clarity. The original version of the draft STSB RAIs are also attached.

Consistent with our communications earlier today, the NRC requests that the responses to the attached final RAIs be issued within 120 days of this email.

Thanks, Perry Buckberg Senior Project Manager phone: (301)415-1383 perry.buckberg@nrc.gov U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Mail Stop O-8B1a Washington, DC, 20555-0001

Hearing Identifier:

NRR_PMDA Email Number:

3754 Mail Envelope Properties (Perry.Buckberg@nrc.gov20171004113600)

Subject:

Request for Additional Information - St. Lucie RICT LAR - MF5372/MF5373 Sent Date:

10/4/2017 11:36:21 AM Received Date:

10/4/2017 11:36:00 AM From:

Buckberg, Perry Created By:

Perry.Buckberg@nrc.gov Recipients:

"Kilby, Gary" <Gary.Kilby@fpl.com>

Tracking Status: None "Snyder, Mike" <Mike.Snyder@fpl.com>

Tracking Status: None "Mack, Jarrett (Jarrett.Mack@fpl.com)" <Jarrett.Mack@fpl.com>

Tracking Status: None "Frehafer, Ken" <Ken.Frehafer@fpl.com>

Tracking Status: None Post Office:

Files Size Date & Time MESSAGE 1460 10/4/2017 11:36:00 AM St Lucie EEOB RAIs RICT August 2017.pdf 129163 St Lucie STSB RAIs RICT August 2017.pdf 131051 St Lucie APLA RICT RAIs October 2017.pdf 155321 DRAFT St Lucie STSB RAIs 505-RICT July 2017.docx 30613 Options Priority:

Standard Return Notification:

No Reply Requested:

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Normal Expiration Date:

Recipients Received:

REQUEST FOR ADDITIONAL INFORMATION ST LUCIE NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-335 AND 50-389 LICENSE AMENDMENT REQUEST TO PERMIT THE USE OF RISK INFORMED COMPLETION TIMES CACs MF5372 & MF5373 By letter dated December 5, 2014, as supplemented by letters dated July 8, 2016, July 22, 2016 and February 25, 2017, Florida Power and Light Company (FPL, the licensee) submitted a license amendment request (LAR) which proposed changes to the Technical Specifications (TSs) for the St. Lucie Nuclear Plant, Units 1 and 2.

Specifically, the requested changes would permit the use of a Risk Informed Completion Time (RICT) for several Technical Specifications. The NRC staff finds that additional information is needed to complete our review of the proposed amendment.

RAI-MF5372/MF5373-EEOB-01 The Commissions Policy on Probabilistic Risk Assessment, Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities, dated August 16, 1995, identifies five key safety principles required for risk-informed decision-making applied to changes to TSs as delineated in RG 1.177 and RG 1.174. They are:

The proposed change meets current regulations; The proposed change is consistent with defense-in-depth philosophy; The proposed change maintains sufficient safety margins; Increases in risk resulting from the proposed change are small and consistent with the Commissions Safety Goal Policy Statement; and The impact of the proposed change is monitored with performance measurement strategies.

NEI 06-09, Risk Informed Technical Specifications Initiative 4b: Risk Managed Technical Specifications (RMTS), Revision 0-A, states that Risk Management Actions (RMAs) and compensatory actions for significant components should be predefined to the extent practicable in plant procedures and implemented at the earliest appropriate time in order to maintain defense-in-depth.

Moreover, the NRC staffs safety evaluation for NEI 06-09, Section 4.0, Limitations and Conditions, (ADAMS No. ML12286A322) states that a licensees LAR adopting the NEI 06-09 initiative will describe the process to identify and provide compensatory measures and RMAs during extended Completion Times (CT), and provide examples of compensatory measures/RMAs.

In Enclosure 12 to the December 5, 2014, LAR, the licensee provided two examples of risk management actions that are considered during a RICT for: a) inoperable diesel generator, and b) inoperable battery.

Provide similar examples of RMAs to assure a reasonable balance of defense-in-depth is maintained for the following TS Actions:

TS LCO/Action Description Current Completion Time Unit 1 3.8.1.1 a One of two offsite circuits inoperable 72 h 3.8.1.1 c One of two offsite A.C. circuits and one of two diesel generator sets inoperable 12 h 3.8.1.1 d Two of the required offsite A.C. circuits inoperable 24 h 3.8.1.1 f One of the Unit 1 startup transformers (1A or 1B) inoperable and a Unit 2 startup transformer (2A or 2B) 72 h 3.8.2.1 Action (undesignated)

Less than required A.C. busses OPERABLE (One or more AC electrical busses inoperable) 8 h Unit 2 3.8.1.1 a One of two offsite circuits inoperable 72 h 3.8.1.1 c One of two offsite A.C. circuits and one of two diesel generator sets inoperable 12 h 3.8.1.1 d Two of the required offsite A.C. circuits inoperable 24 h 3.8.1.1 f One of the Unit 2 startup transformers (2A or 2B) inoperable and a Unit 1 startup transformer (1A or 1B) 72 h 3.8.3.1 a One of the required trains of A.C. Emergency busses not fully energized 8 h 3.8.3.1 b (1)

One A. C. Instrument Bus either not energized from its associated inverter, or with the inverter not connected to its associated D.C. Bus

((1) re-energize the A.C. Instrument Bus) 2 h 3.8.3.1 b (2)

One A. C. Instrument Bus either not energized from its associated inverter, or with the inverter not connected to its associated D.C. Bus

((2) re-energize the A. C. Instrument Bus from its associated inverter connected to its associated D.C. Bus) 24 h 3.8.3.1 c One D.C. bus not energized from its associated battery bank 2 h RAI-MF5372/MF5373-EEOB-02

In the LAR, Enclosure 1, Table E1-1, In Scope TS/LCO Conditions to Corresponding PRA Functions, describes the design success criteria for each TS Limiting Condition for Operation (LCO).

Provide a revised Table E1-1 for the Electrical Power Systems TSs that includes details for each Action statement (see table in RAI-MF5372/MF5373-EEOB-01 depicting details by TS action rather than by LCO) to be utilized in the RICT Program. Provide the design success criteria for each Action and clarify the absolute minimum set of equipment needed to accomplish the safety function.

RAI-MF5372/MF5373-EEOB-03 In the February 25, 2017, LAR supplement, the licensee proposed (for each Unit) to move a NOTE in LCO 3.8.1.1, A.C. Sources, applicable to Actions b and c. The NOTE states, If the absence of any common-cause failure cannot be confirmed, this test shall be completed regardless of when the inoperable EDG is restored to OPERABILITY, and is indicated by an asterisk that it applies to the statement, demonstrate the OPERABILITY of the remaining OPERABLE EDG by performing Surveillance Requirement 4.8.1.1.2.a.4 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months unless it can be confirmed that the cause of the inoperable EDG does not exist on the remaining EDG.

When the NOTE for Action b and Action c are re-positioned above each Action and the asterisks are deleted, the applicability of this test is no longer clear, as there are two Surveillance Requirements in both Action statements. Provide a revised TS Markup for 3.8.1.1 for St. Lucie Units 1 and 2 to clarify these discrepancies.

RAI-MF5372/MF5373-EEOB-04 In Enclosure 1 of the LAR, Tables E1-2, Unit 1 In Scope TS/LCO Conditions RICT Estimate, and E1-3, Unit 2 In Scope TS/LCO Conditions RICT Estimate, detail the licensees estimated RICT calculations for Units 1 and 2, respectively. The licensee provided estimates of less than (<) 1 day for the following Electrical Power System TS Actions:

TS LCO/Action Description Current Completion Time Unit 1 3.8.2.1 Action (undesignated)

Less than required A.C. busses OPERABLE (One or more AC electrical busses inoperable) 8 h 3.8.2.3 a One of the required battery banks or busses Inoperable 2 h Unit 2 3.8.2.1 a One battery bank inoperable 2 h 3.8.3.1 a One of the required trains of A.C. Emergency busses not fully energized 8 h

3.8.3.1 c One D.C. bus not energized from its associated battery bank 2 h For the above Actions, provide the estimated RICT values in hours.

RAI-MF5372/MF5373-EEOB-05 St. Lucie's Updated Final Safety Analysis Report, Chapter 8, states:

The onsite ac and dc power systems are designed with redundancy and independence of onsite power sources, buses, switchgear, distribution cabling and controls to provide reliable supply of electrical power to safety related electrical loads needed to achieve safe plant shutdown or to mitigate the consequences of a design basis accident.

When the licensee enters the TS 3.8.2.1 Action (undesignated), these buses carry the potential vulnerability to single failures that will reduce protection against the exceedance of the design limits.

TS LCO/Action Description Current Completion Time Unit 1 3.8.2.1 Action (undesignated)

Less than the complement of A.C. busses OPERABLE (One or more AC electrical busses inoperable) 8 h

1) For the above TS condition's lowest estimated RICT (least amount of time available, calculated beyond the front-stop):

a. Describe a scenario/plant configuration for this condition.

b. Explain how each bus would retain the ability to defend against vulnerabilities during this scenario (e.g., examples of RMAs to assure a reasonable balance of defense-in-depth is maintained for this TS condition).

2) For the above TS condition's highest estimated RICT (most risk significant component(s) that would result in a calculation close to the 30-day back-stop, without Probabilistic Risk Assessment (PRA) functional consideration):

a. Describe a scenario/plant configuration for this condition.

b. Explain how each bus would retain the ability to defend against vulnerabilities during this scenario (e.g., examples of RMAs to assure a reasonable balance of defense-in-depth is maintained for this TS condition).

REQUEST FOR ADDITIONAL INFORMATION ST LUCIE NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-335 AND 50-389 LICENSE AMENDMENT REQUEST TO PERMIT THE USE OF RISK INFORMED COMPLETION TIMES CACs MF5372 & MF5373 By letter dated December 5, 2014, as supplemented by letters dated July 8, 2016, July 22, 2016 and February 25, 2017, Florida Power and Light Company (FPL, the licensee) proposed changes to the Technical Specifications (TSs) for the St. Lucie Nuclear Plant, Units 1 and 2. Specifically, the requested changes would permit the use of a Risk Informed Completion Time (RICT) for several Technical Specifications. The NRC staff finds that additional information is needed to complete our review of the proposed amendment.

In the letter dated February 25, 2017, FPL stated that Attachments 3 and 4 to the letter provide a complete markup of the TS for this LAR, and supersede the TS markups provided previously. The letter also states FPLs intended approach in this supplement is to remove loss of function provisions.

The categories of items required to be in the TSs are provided in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(c). As required by 10 CFR 50.36(c)(2)(i),

the TSs will include limiting conditions for operation (LCOs), which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. Per 10 CFR 50.36(c)(2)(i), when an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.

Within the context of the RICT program, a TS Loss of Function (TS LOF) is considered to exist when there is insufficient OPERABLE equipment to fulfill a safety function.

Additional administrative controls are needed to support the application of a RICT to TS LOF conditions due to safety margin and defense-in-depth considerations.

The staff requests the following information to support a determination that the proposed remedial actions and time frames for completion are appropriate.

RAI-MF5372/MF5373-STSB-01:

Unit 1 TS LCO 3.4.3 and Unit 2 TS LCO 3.4.2.2 require that all pressurizer code safety valves shall be OPERABLE with specified lift settings. ACTION a is applicable when one pressurizer code safety valve is inoperable, and requires, in part, restoring the inoperable valve to Operable status within 15 minutes. The LAR proposes to apply a RICT to this ACTION.

The TS Bases state that during operation, all pressurizer Code safety valves must be OPERABLE to prevent the Reactor Coolant System from being pressurized above its Safety Limit. Based on this statement, it appears that the safety function could not be accomplished if one Code safety valve is inoperable.

If this condition does not represent a loss of function, please provide technical justification supporting that conclusion. If this condition does represent a loss of function, please provide additional justification for its inclusion in the RICT program, including appropriate administrative controls and explain how safety margin and defense-in-depth considerations are maintained.

RAI-MF5372/MF5373-STSB-02:

Unit 1 TS LCO 3.4.12 and Unit 2 TS 3.4.4 require that each Power Operated Relief Valve (PORV) Block Valve shall be OPERABLE. ACTION a applies when one or more block valves are inoperable, and requires restoration of the block valve(s) to OPERABLE status within one hour or close the block valve(s) and remove power from the block valve(s).

Unit 1 UFSAR Section 5.5.3.2 describes the PORVs as half-capacity valves with a motor actuated isolation valve upstream of each of the PORVs to permit isolating the (PORV) valve for maintenance or in case of valve failure.

With both block valves closed and de-energized, operation of the PORVs could be delayed until power could be restored to the block valves. Please provide a description of the actions and general time frames that would be required to re-energize the block valves so that the PORVs could be opened. Please provide a summary of the UFSAR Chapter 15 analyses in which operation of the PORVs is credited; and explain why delays in PORV operation are consistent with the analyses presented.

RAI-MF5372/MF5373-STSB-03:

For Units 1 and 2, TS LCO 3.5.1 requires that each Reactor Coolant System safety injection tank shall be OPERABLE.

ACTION b applies with one safety injection tank inoperable, except as a result of parameter limits specified in ACTION a, not being within limits. The LAR proposes to apply a RICT to this ACTION.

Unit 1 UFSAR Section 6.2.1.3.2, Containment Vessel Transient Analysis, and Unit 2 UFSAR Section 6.2.1.1.3, Design Evaluation - Containment Pressure - Temperature Analysis, state that the LOCA accident analyses are based upon the following additional overall assumptions:

For the discharge leg break, the contents of three safety injection tanks (SITs) discharge into the reactor vessel when reactor coolant system pressure drops below tank pressure. This assumes the entire contents of the safety injection tank in the ruptured leg does not reach the core. For the hot and suction leg cases the contents of four SITs is considered.

It is not clear to the staff how the assumptions in the accident analysis would be satisfied for a LOCA in which the contents of one accumulator is lost through the break, and a second accumulator is inoperable at the time of the event.

If this condition does not represent a loss of function, please provide technical justification supporting that conclusion. If this condition does represent a loss of function, please provide additional justification for its inclusion in the RICT program, including appropriate administrative controls and explain how safety margin and defense-in-depth considerations are maintained.

RAI-MF5372/MF5373-STSB-04:

For Units 1 and 2, TS LCO 3.6.2.1 requires that two containment spray trains and two containment cooling trains be operable. ACTION e applies when two containment cooling trains are inoperable, and requires, in part, that one cooling train be restored to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The LAR proposes to include this ACTION in the scope of the RICT program.

Unit 1 UFSAR Section 6.2.1.3.2, Containment Vessel Transient Analysis, and Unit 2 UFSAR Section 6.2.1.1.3, Design Evaluation - Containment Pressure - Temperature Analysis, provide a summary of the containment peak pressure and temperature analysis. For the LOCA, the listing of input assumptions includes the following:

The analyses are based on the loss of offsite power in which a coincident loss of diesel generator is assumed. This results in the loss of one cooling train which disables two fan coolers and one containment spray. This leaves one containment spray pump and one train of fan coolers (i.e. two units) available for operation.

It is not clear to the staff how the assumptions in the accident analysis would be satisfied if two containment cooling trains are inoperable.

If this condition does not represent a loss of function, please provide technical justification supporting that conclusion. If this condition does represent a loss of function, please provide additional justification for its inclusion in the RICT program, including appropriate administrative controls and explain how safety margin and defense-in-depth considerations are maintained.

RAI-MF5372/MF5373-STSB-05:

Unit 1 TS LCO 3.6.3.1 and Unit 2 TS LCO 3.6.3 require that the containment isolation valves be operable. The Actions are applicable with one or more of the isolation valve(s) inoperable and require, in part, restoration of the valve(s) to operable status or isolation of each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The LAR proposes to include these Actions in the scope of the RICT program.

It is not clear to the staff how the assumptions regarding containment isolation would be satisfied if more than one valve in a given penetration is inoperable.

If this condition does not represent a loss of function, please provide technical justification supporting that conclusion. If this condition does represent a loss of

function, please provide additional justification for its inclusion in the RICT program, including appropriate administrative controls and explain how safety margin and defense-in-depth considerations are maintained.

RAI-MF5372/MF5373-STSB-06:

For Unit 2, TS LCO 3.6.1.7 requires that each containment purge supply and exhaust valve be OPERABLE with each 48-inch containment purge supply and exhaust isolation valve sealed closed and the 8-inch containment purge supply and exhaust isolation valves be open for purging and/or venting as required for safety related purposes such as those listed in the LCO.

ACTION a applies with a 48-inch containment purge supply and/or exhaust isolation valve(s) open for reasons other than maintaining containment pressure or reducing containment atmosphere airborne radioactivity and/or improving air quality. The ACTION requires closing the valve or isolating the penetration(s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The proposed change is to allow the calculation of a RICT for this configuration, which would allow postponement of isolating the penetration for up to 30 days.

ACTION b applies with an 8-inch containment purge supply and/or exhaust isolation valve(s) open for reasons other than maintaining containment pressure or reducing containment atmosphere airborne radioactivity and/or improving air quality. The ACTION requires closing the valve or isolating the penetration(s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The proposed change is to allow the calculation of a RICT for this configuration, which would allow postponement of isolating the penetration for up to 30 days.

ACTION c applies with containment purge supply and/or exhaust isolation valve(s) having a measured leakage rate exceeding the limits specified in the Surveillance Requirements. The ACTION requires, in part, restoring the valve to operable status or isolating the penetrations such that the measured leakage rate does not exceed the limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The proposed change is to allow the calculation of a RICT for this configuration, which would allow postponement of restoring the valve to operable status or isolating the penetration for up to 30 days.

The St. Lucie Unit 2 UFSAR Section 15.6.5.5.2, Compliance with RG 1.183 Regulatory Positions, Compliance with Regulatory Position 3.8 states:

100% of the radionuclide inventory of the RCS is released instantaneously at the beginning of the event. The containment purge flow is 2500 cfm through the eight-inch line and is assumed to be isolated after 30 seconds. No filters are credited.

Please explain how the specified safety function of the containment purge portion of the containment ventilation system would be accomplished during application of a RICT to these ACTIONS. Please explain how the proposed changes would ensure the assumptions regarding isolation of the containment purge system in the accident analysis are satisfied.

RAI-MF5372/MF5373-STSB-07:

For Units 1 and 2, TS LCO 3.7.1.5 requires that each main steam line isolation valve (MSIV) be Operable. The ACTION for Mode 1 requires, in part, that with one MSIV inoperable, Power Operation may continue provided the inoperable valve is restored to Operable status or closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The proposed change is to allow the calculation of a RICT for this configuration, which would allow postponement of restoring the valve to operable status for up to 30 days.

The TS Bases state that the operability of the MSIVs ensures that no more than one steam generator will blow down in the event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment.

It is not clear to the staff how the assumptions in the accident analysis would be satisfied when one MSIV is inoperable.

If this condition does not represent a loss of function, please provide technical justification supporting that conclusion. If this condition does represent a loss of function, please provide additional justification for its inclusion in the RICT program, including appropriate administrative controls and explain how safety margin and defense-in-depth considerations are maintained.

RAI-MF5372/MF5373-STSB-08:

For Units 1 and 2, TS LCO 3.7.1.5 requires that each main steam line isolation valve (MSIV) be Operable.

The existing ACTION for Mode 1 states:

With one main steam line isolation valve inoperable, POWER OPERATION my continue provided the inoperable valve is either restored to OPERABLE status or closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; otherwise be in HOT STANDBY with the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The existing ACTION for Modes 2 and 3 states:

With one or both main steam isolation valve(s) inoperable, subsequent operation in MODES 2 or 3 may proceed provided the isolation valve(s) is (are) maintained closed. Otherwise, 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 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The proposed ACTION for Mode 1 would state:

With one main steam line isolation valve inoperable, POWER OPERATION my continue provided the inoperable valve is either restored to OPERABLE status or closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or in accordance with the RICT Program; otherwise be in MODE 2 with the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The proposed ACTION for Modes 2 and 3 would state:

With one or both main steam isolation valve(s) inoperable, subsequent operation in MODES 2 or 3 may continue provided:

1. The inoperable main steam isolation valves are closed within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and

2. The inoperable main steam isolation valves are verified closed once per 7 days.

Otherwise, 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 HOT SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The proposed changes would align the St. Lucie LCO Actions with the corresponding Actions in the Standard Technical Specifications. In the Standard Technical Specifications, the time allowed to close an inoperable main steam isolation valve while in MODES 2 or 3 is a bracketed value. Values enclosed in brackets are used to signify a licensee-specific value. Please provide the technical justification for the selection of the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months value for St. Lucie.

RAI-MF5372/MF5373-STSB-09:

TS 6.8.4 states, The following program shall be established, implemented, and maintained. Proposed TS 6.8.4.q describes the RICT program.

Element c requires that when a RICT is being used, any plant configuration change within the scope of the RICT program must be considered for the effect on the RICT.

The terminology within the scope of the RICT program could be misinterpreted to only include equipment governed by TS LCOs that are included within the RICT program. In accordance with NEI 06-09, any plant configuration change, as defined in NEI 06-09 0-A, must be considered for its effect on the RICT. Please propose revised language to reflect the broader scope of changes that could affect the RICT.

The proposed TS 6.8.4.q does not address the treatment of common cause when an emergent failure occurs. Please propose additional administrative controls to address the treatment of common cause.

REQUEST FOR ADDITIONAL INFORMATION ST LUCIE NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-335 AND 50-389 LICENSE AMENDMENT REQUEST TO PERMIT THE USE OF RISK INFORMED COMPLETION TIMES CACs MF5372 & MF5373 By letter dated December 5, 2014, as supplemented by letters dated July 8, 2016, July 22, 2016 and February 25, 2017, Florida Power and Light Company (FPL, the licensee) submitted a license amendment request (LAR) which proposed changes to the Technical Specifications (TSs) for the St. Lucie Nuclear Plant, Units 1 and 2. Specifically, the requested changes would permit the use of a Risk Informed Completion Time (RICT) for several Technical Specifications.

The NRC staff finds that additional information is needed to complete our review of the proposed amendment.

RAI-MF5372/73-APLA-01.c.R1 (Internal event probabilistic risk assessment (PRA))

The July 8, 2016, response to RAI-MF5372/73-APLA-01.c (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16193A659) states that the initiating event frequency associated with loss of a single 120 V instrument bus is 9E-04/year and states that the common cause failure of 120 V instrument buses would only have a very small impact on RICT estimates. It is not clear why the common cause failure (CCF) of 120 V instrument buses would only have a very small impact on RICT estimates, given that failure of redundant 120 V instrument buses could lead to an unanticipated system response. The NRC staff notes that although the CCF frequency of 120 V instrument buses would be less than the failure frequency of a single bus, the criteria for screening initiating events based on Supporting Requirement IE-C6 of the Probabilistic Risk Assessment (PRA) Standard would likely not be met (i.e., an initiating event can be screened if the frequency of the event is less than 1E-06/year and the core damage would not occur unless at least two trains of mitigating systems are failed independent of the initiator).

a) Summarize the anticipated plant responses to the common mode failure of redundant 120 V instrument buses.

b) Based on the plant response, justify the exclusion of the CCF of redundant 120 V instrument buses as an initiating event contributor.

c) Provide an explanation on how excluding CCF events for these buses affects the CCF calculations described in RAIs RAI-MF5372/73 12 and 13 below.

RAI-MF5372/73-APLA-02.b.R1 (Fire PRA)

The response to RAI-MF5372/73-APLA-02 part b explains that the PRA model and update process required by the PRA procedure ensures the model is consistent with the as-built, as-operated plant by using various triggering factors. The response does not describe the factors that will be considered and monitored to trigger an update, such as plant modifications, procedure changes, changes in industry information and other changes. Accordingly, it is not clear to the NRC staff how the procedure ensures that the models used in the RICT program will reflect the as-built, as-operated plant. Describe the factors considered and monitored that

would trigger an update such as plant modifications, procedure changes, changes in industry information and other changes to ensure the models used in the RICT program reflect the as-built, as-operated plant.

RAI-MF5372/73-APLA-09.R1 (NFPA 805 Modification Implementation)

The response to RAI-MF5372/73-APLA-09 states that at the time of implementation of [a]

RICT, any modifications that are not installed will not be credited in the estimation of [core damage frequency] or [large, early release frequency]. Using this approach, it is possible, depending on the modifications that remain to be installed at the time of a RICT implementation, that the as-built, as-operated plant may not meet Regulatory Guide (RG) 1.174 risk acceptance guidelines. Explain when the modification that will place the risk below the RG 1.174 acceptance guidelines is expected to be completed and what the risk is expected to be. Also, provide a license condition that will verify that the as-built, as-operated plant meets RG 1.174 risk acceptance guidelines at the time a RICT is implemented.

RAI-MF5372/73-APLA-09.R1b (NFPA 805 Modification Implementation)

The response to RAI-MF5372/73-APLA-09 states that at the time of implementation of [a]

RICT, any modifications that are not installed will not be credited in the estimation of CDF or LERF. Using this approach it is possible, depending on the modifications that remain to be installed at the time of a RICT implementation, that the as-built, as-operated plant may not meet RG 1.174 risk acceptance guidelines.

As discussed in the March 31, 2016, safety evaluation for the amendment to transition the fire protection program to Section 50.48(c) of Title 10 of the Code of Federal Regulations (10 CFR)

(ADAMS Accession No. ML15344A346), FPL used the guidance in frequently asked question (FAQ) 08-0046, "Closure of National Fire Protection Association 805 Frequently Asked Question 08-0046 Incipient Fire Detection Systems," to incorporate its very early warning fire detection system (VEWFDS) into the fire PRA. In December of 2016, the NRC staff published NUREG-2180, "Determining the Effectiveness, Limitations, and Operator Response for Very Early Warning Fire Detection Systems in Nuclear Facilities, (Delores-VEWFIRE)," which included new guidance on modeling VEWFDS. The methodology in NUREG-2180 is currently the best available guidance and replaces the guidance in FAQ 08-0046, which has been retired.

By letter dated November 17, 2016 (ADAMS Accession No. ML16253A111), the NRC staff informed the industry that, [i]f a licensee is performing a periodic or interim PRA update, performing a fire risk evaluation in support of self-approval, or submitting a future risk informed license amendment request, the staffs expectation is that they will assess the impact of new operating experience and information [e.g., NUREG-2180] on their PRA analyses and incorporate the change as appropriate per Regulatory Guide 1.200, Revision 2.

a) If FPL will use the methodology in NUREG-2180, please provide:

1. An estimate of the current CDF and LERF for all quantified hazards using the NUREG-2180 methodology in the fire PRA.

2. If the current CDF and LERF estimates do not satisfy the limitations and conditions in Section 4, item 6 of the NEI 06-09 safety evaluation, provide an explanation of how these guidelines will be met before implementation of the RICT program.

3. If the methodology (e.g., approach, methods, data, and assumptions) has not been incorporated into the fire PRA (i.e., PRA model changes and documentation completed and the upgrade peer reviewed), provide an explanation of when it will be incorporated into the PRA model of record that will be used to estimate RICTs (the response may reference the response to RAI-MF5372/73-APLA-16 (below) which requests a list of implementation items).

b) If FPL proposes not to use the methodology in NUREG-2180, please provide:

1. Confirmation that the methodology in the retired FAQ 08-0046 is not the proposed methodology.

2. A description of the proposed methodology (e.g., approach, methods, data, and assumptions) that will be used in the fire PRA. The description should include a detailed comparison of that proposed methodology with the methodology in NUREG-2180.

3. Justification of the proposed methodology including comparison with available experimental results. Development and use of a proposed alternative may result in additional RAIs and significantly extend the time and resources required to complete the review.

4. An estimate of the current CDF and LERF for each quantified hazard with fire PRA results: (1) without credit for VEWFDS; (2) that would be obtained had the guidance in NUREG-2180 been applied, and (3) obtained using the proposed methodology.

5. If the current CDF and LERF estimates do not satisfy the limitations and conditions in Section 4, item 6 of the NEI 06-09 safety evaluation, provide an explanation of how these guidelines will be met before implementation of the RICT program.

6. An evaluation on how using the proposed methodology instead of the NUREG-2180 methodology could impact the RICT estimates.

7. If the methodology (e.g., approach, methods, data, and assumptions) has not been incorporated into the fire PRA (i.e., PRA model changes and documentation completed and the upgrade peer reviewed), provide an explanation of when it will be incorporated into the PRA model of record that will be used to calculate the RICTs (response may reference the response to RAI-MF5372/73-APLA-16 which requests a list of implementation items).

RAI-MF5372/73-APLA-11.R1 Instrumentation Models In response to RAI-MF5372/73-APLA-11, regarding the level of detail in the modeling of TS 3.3.2.1 (Unit 1) and TS 3.3.2 (Unit 2) for Engineered Safety Features Actuation System Instrumentation, the licensee stated that the PRA model includes the individual instrumentation channels; therefore, inoperability of individual instrument channels can be assessed directly by the PRA for the RICT Program.

a) Please explain how instrumentation is modelled in the PRA. If there are different types of models (e.g., multiple channel basic events versus a single combined basic event) that are used for different instrumentation, please explain all the different models.

b) Clarify how each of the models will be changed to model the impact of an unavailable channel and why this modelling given one unavailable channel is correct or will conservatively bound the RICT calculation.

c) If any instrument channels included by the Risk Managed Technical Specifications (RMTS) program are not modeled at all in the PRA, explain how that instrument channel failure will be included in the RICT estimate.

RAI-MF5372/73-APLA-12 Clarification of Less than Design Basis Capability Reductions in the functional capability below the design basis success criteria and vulnerable plant configurations with currently short completion times, will reduce the defense in depth and safety margins available below previously accepted levels while the plant operates in that condition during an extended completion time. The NRC staffs safety evaluation for Nuclear Energy Institute (NEI) 06-09, Risk-Informed Technical Specifications Initiative 4b - Risk-Managed Technical Specifications (RMTS) Guidelines, as well as the constraints regarding PRA Functionality, limit the use of PRA Functional by specifying that the remaining function capability continues to meet the design basis success criteria parameters (e.g. maintaining the functional capability to perform at the level of one operable train) unless appropriate disposition and restrictions are provided. For example, in the submittal the licensee proposes an LCO condition associated with the Code Safety Valves that appears to have a RICT calculation, even though the deign bases success criteria parameter values do not seem to be capable of being met which may be a TS Loss of Function. Table E1-1, In Scope TS/LCO Conditions to Corresponding PRA Functions, of the December 5, 2014, LAR indicates that the design basis success criteria is 3 of 3 valves while the PRA success criteria is 2 of 3 valves. Also, there may be other TS loss of function conditions which, if entering a RICT, also are not capable of meeting the design basis success criteria.

For each loss of function LCO condition that includes a RICT, but where the design basis success criteria parameters may not be modelled or may not be met for the PRA success criteria:

a) Identify the design basis parameters that may not be available (e.g., the relief capability credited in the design basis that requires 3 of 3 versus 2 of 3 valves).

b) Identify the design basis accident scenarios that rely on those parameters (e.g., what design basis accidents require greater than 2 valve pressure relief capabilities).

c) For each of these accident scenarios, explain the impact of only having the PRA parameter capabilities on the affected design basis success accidents. Include a justification of the effect this change in available capabilities will have on defense-in-depth and safety margins.

d) If new PRA functional parameters are proposed, identify how these parameters will be included in the technical specifications.

e) If new PRA functional parameters are not proposed, remove the conditions from the program.

RAI-MF5372/73-APLA-13 Common Cause Failure Terms for Planned Maintenance While the guidance in NEI 06-09 states that no CCF adjustment is required for planned maintenance, the NRC approval of NEI 06-09 is based on RG 1.177, as indicated in the NRC safety evaluation to NEI 06-09 (ADAMS Accession No. ML071200238). Specifically, Section 2.2 of the NRC staffs safety evaluation for NEI 06-09 states that, specific methods and guidelines acceptable to the NRC staff are [] outlined in RG 1.177 for assessing risk-informed TS changes. Further, Section 3.2 of the NRC staffs safety evaluation states that compliance with the guidance of RG 1.174 and RG 1.177, is achieved by evaluation using a comprehensive risk analysis, which assesses the configuration-specific risk by including

contributions from human errors and common cause failures.

The guidance in RG 1.177, Section 2.3.3.1, states that, CCF modeling of components is not only dependent on the number of remaining inservice components, but is also dependent on the reason components were removed from service (i.e. whether for preventative or corrective maintenance). In relation to CCF for preventive maintenance, the guidance in RG 1.177, Appendix A, Section A-1.3.1.1, states:

If the component is down because it is being brought down for maintenance, the CCF contributions involving the component should be modified to remove the component and to only include failures of the remaining components (also see Regulatory Position 2.3.1 of Regulatory Guide 1.177).

According to RG 1.177, if a component from a CCF group of three or more components is declared inoperable, the CCF of the remaining components should be modified to reflect the reduced number of available components in order to properly model the as-operated plant.

a)

Please explain how CCF are included in the PRA model (e.g., with all combinations in the logic models as different basic events or with identification of multiple basic events in the cut sets) b)

Please explain how the quantification and/or models will be changed when, for example, one train of a 3X100 percent train system is removed for preventative maintenance and describe how the treatment of CCF either meets the guidance in RG 1.177 or meets the intent of this guidance when quantifying a RICT.

RAI-MF5372/73-APLA-14 Evaluation of Common Cause for Emergent Failures According to Section A-1.3.2.1 of Appendix A of RG 1.177, when a component fails, the CCF probability for the remaining redundant components should be increased to represent the conditional failure probability due to CCF of these components, in order to account for the possibility that the first failure was caused by a CCF mechanism. When a component fails, the calculation of the plant risk, assuming that there is no increase in CCF potential in the redundant components before any extent of condition evaluation is completed, could lead to a non-conservative extended completion time calculation, as illustrated by inclusion of the guidance in Appendix A of RG 1.177. Much of the discussion in Appendix A describes how configuration specific risk calculations should be performed.

In Section 3.2 of the NRC staffs safety evaluation for NEI 06-09, the staff states that compliance with the guidance of RG 1.174 and RG 1.177, is achieved by evaluation using a comprehensive risk analysis, which assesses the configuration-specific risk by including contributions from human errors and common cause failures.

The requirement to consider additional risk mitigation actions (RMAs) prior to the completion of the extent of cause evaluation was included by the NRC staff in the safety evaluation for NEI 06-09 as an additional measure to account for the increased potential that the first failure was caused by a CCF mechanism. However, no exception to the RG 1.177 guidance was taken in the calculation of the RICT with regards to the quantification of the unresolved potential for CCF before the extent of cause evaluation is complete. The NRC staff interprets that the combined guidance in RG 1.177 and NEI 06-09 0-A could be met with the following process.

When, prior to exceeding the front stop, there is a high degree of confidence based on the evidence collected that there is no common cause failure mechanism that could affect the redundant components, the RICT calculation may use nominal Common Cause factor probability.

If a high degree of confidence cannot be established that there is no common cause failure that could affect the redundant components, the RICT shall account for the increased possibility of common cause failure. Accounting for the increased possibility of common cause failure shall be accomplished by one of the two methods below. If one of the two methods below is not used, the TS front stop shall not be exceeded.

1. The RICT calculation shall be adjusted to numerically account for the increased possibility of common cause failure, in accordance with RG 1.177, as specified in Section A-1.3.2.1 of Appendix A of the RG. Specifically, when a component fails, the common cause failure probability for the remaining redundant components shall be increased to represent the conditional failure probability due to common cause failure of these components, in order to account for the possibility the first failure was caused by a common cause mechanism.

2. Prior to exceeding the front stop, the licensee shall implement RMAs in addition to those already credited in the RICT calculation, that target the success of the redundant and/or diverse structures, systems or components (SSC) of the failed SSC, and, if possible, reduce the frequency of initiating events which call upon the function(s) performed by the failed SSC. Documentation of RMAs shall be available for NRC review.

a) Please confirm and describe how that treatment of CCF, in the case of an emergent failure, either meets the guidance in RG 1.177 or meets the intent of this guidance together with the NEI 06-09 0-A guidance when quantifying a RICT.

b) Please propose where the guidance on how CCFs will be treated will be placed to ensure that the guidance is followed, e.g., as a license condition or in the Administrative TS that implements the RICT program.

RAI-MF5372/73-APLA-15 License Condition In Section 4.0, "Limitations and Conditions" of the NRC staffs safety evaluation for NEI 06-09, the staff states:

As part of its review and approval of a licensee's application requesting to implement the RMTS, the NRC staff intends to impose a license condition that will explicitly address the scope of the PRA and non-PRA methods approved by the NRC staff for use in the plant specific RMTS program. If a licensee wishes to change its methods, and the change is outside the bounds of the license condition, the licensee will need NRC approval, via a license amendment, of the implementation of the new method in its RMTS program.

Please propose a license condition limiting the scope of the PRA and non-PRA methods to what is approved by the NRC staff for use in the plant-specific RMTS program. Wording consistent with the example below would be acceptable.

The risk assessment approach, methods, and data shall be acceptable to the NRC, be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk from extending the completion times must be PRA methods accepted as part of this license amendment, or other methods currently approved by the NRC for generic use. If a licensee wishes to change its methods and the change is outside the bounds of this license condition, the licensee will need prior NRC approval, via a license amendment.

RAI-MF5372/73-APLA-16 Implementation Items Please provide a list of activities (i.e., implementation items) that are credited as part of the approval of the request to implement a RICT program that will not be completed before issuing the amendment but must be complete prior to implementation of the RICT program.

RAI-MF5372/73-APLA-17 External Events NEI 06-09, Section 3.3.5, External Events Consideration clarifies that external hazards impact on incremental configuration risk should be addressed for each RICT calculation. Enclosure 4 of the December 5, 2014, LAR, Information Supporting Justification of Excluding Sources of Risk not addressed by the PRA Models, addresses external events. The Enclosure summarizes the evaluation of the risk of external hazards that appears to be consistent with the ASME/ANS PRA Standard, i.e., screening associated with the baseline risk contribution. The results of the evaluation summarized in LAR Table E4-1, Evaluation of Risks from External Hazards, seem to indicate, however, that all external hazards will be excluded from every configuration risk evaluation, e.g. no unique PRA model for seismic events is required in order to assess configuration risk for the RICT Program. However, there may be situations where the hazard may be important in a configuration risk calculation even though the baseline risk can be screened out consistent with the ASME/ANS PRA Standard. For example, external floods seem to be excluded because the plant design conforms to the Standard Review Plan (SRP) criteria. Presumably, smaller flood levels may fail plant equipment not required to be protected by the SRP criteria which could affect configuration risk, and sometimes the flood barriers themselves may be degraded or undergoing maintenance which could affect configuration risk. Similarly, extreme wind seems to be fully excluded because of low frequency of occurrence and SRP conformance, but these factors may not have considered the plant configuration during a RICT.

Please clarify if all external hazard risks are excluded from the RICT program or if the program includes guidance to assure that the assumptions supporting the screening of the hazards remain applicable given the plant configuration during the RICT. If all hazards are fully excluded, please address the issue related to screening based on meeting the SRP criteria (e.g., design flood height and mitigating features) or based on low nominal risk values. If, instead, guidance is provided, please describe the guidance (e.g., in certain instances, hazards which were initially screened out from the RICT calculation may be considered quantitatively if the plant configuration could impact the RICT).

RAI-MF5372/73-APLA-18 Dual Unit Impacts

The proposed changes to the TSs sometimes differ between Unit 1 and Unit 2 (e.g., 3.6.1.7, page 11 of 23), and sometimes do not. The RICT estimates also appear to sometimes differ between the units (e.g., 3.6.2.1.a), and sometimes do not.

a)

Clarify whether there are two independent baseline PRAs and, if not, how are the unit specific risk estimates developed?

b)

Clarify whether two independent CRMP models will be developed and used in parallel and, if not, how will unit specific RICTs be estimated?

c)

Explain how the unit-specific RICT calculations will be appropriately estimated given any differences between the units, interactions between the units caused by the shared equipment, and the different out of service SSCs in each unit.

DRAFT REQUEST FOR ADDITIONAL INFORMATION ST LUCIE NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-335 AND 50-389 LICENSE AMENDMENT REQUEST TO PERMIT THE USE OF RISK INFORMED COMPLETION TIMES CACs MF5372 & MF5373 By letter dated December 5, 2014, as supplemented by letters dated July 8, 2016, July 22, 2016 and February 25, 2017, Florida Power and Light Company (FPL, the licensee) proposed changes to the Technical Specifications (TSs) for the St. Lucie Nuclear Plant, Units 1 and 2. Specifically, the requested changes would permit the use of a Risk Informed Completion Time (RICT) for several Technical Specifications. The NRC staff finds that additional information is needed to complete our review of the proposed amendment.

In the letter dated February 25, 2017, FPL stated that Attachments 3 and 4 to the letter provide a complete markup of the TS for this LAR, and superseded the TS markups provided previously. The letter also states FPLs intended approach in this supplement is to remove loss of function provisions.

The categories of items required to be in the TSs are provided in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(c). As required by 10 CFR 50.36(c)(2)(i),

the TSs will include limiting conditions for operation (LCOs), which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. Per 10 CFR 50.36(c)(2)(i), when an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.

St. Lucie TS 5.5.15 describes a loss of function as:

A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed.

Within the context of the RICT program, a TS Loss of Function (TS LOF) is considered to exist when there is insufficient OPERABLE equipment to fulfill a safety function.

Additional administrative controls are needed to support the application of a RICT to TS LOF conditions due to safety margin and defense-in-depth considerations.

The staff requests the following information to support a determination that the proposed remedial actions and time frames for completion are appropriate.

RAI-MF5372/MF5373-STSB-01:

Unit 1 TS LCO 3.4.3 and Unit 2 TS LCO 3.4.2.2 require that all pressurizer code safety valves shall be OPERABLE with specified lift settings. ACTION a is applicable when one pressurizer code safety valve is inoperable, and requires, in part, restoring the

inoperable valve to Operable status within 15 minutes. The LAR proposes to apply a RICT to this ACTION.

The TS Bases state that during operation, all pressurizer Code safety valves must be OPERABLE to prevent the Reactor Coolant System from being pressurized above its Safety Limit. Based on this statement, it appears that the safety function could not be accomplished if one Code safety valve is inoperable.

If this condition does not represent a loss of function, please provide technical justification supporting that conclusion. If this condition does represent a loss of function, please provide additional justification for its inclusion in the RICT program, including appropriate administrative controls and explain how safety margin and defense-in-depth considerations are maintained.

RAI-MF5372/MF5373-STSB-02:

Unit 1 TS LCO 3.4.12 and Unit 2 TS 3.4.4 require that each Power Operated Relief Valve (PORV) Block Valve shall be OPERABLE. ACTION a applies when one or more block valves are inoperable, and requires restoration of the block valve(s) to OPERABLE status within one hour or close the block valve(s) and remove power from the block valve(s).

Unit 1 UFSAR Section Section 5.5.3.2 describes the PORVs as half-capacity valves with a motor actuated isolation valve upstream of each of the PORVs to permit isolating the (PORV) valve for maintenance or in case of valve failure.

With both block valves closed and de-energized, operation of the PORVs could be delayed until power could be restored to the block valves. Please provide a description of the actions and general time frames that would be required to re-energize the block valves so that the PORVs could be opened. Please provide a summary of the UFSAR Chapter 15 analyses in which operation of the PORVs is credited; and explain why delays in PORV operation are consistent with the analyses presented.

RAI-MF5372/MF5373-STSB-03:

For Units 1 and 2, TS LCO 3.5.1 requires that each Reactor Coolant System safety injection tank shall be OPERABLE.

ACTION b applies with one safety injection tank inoperable, except as a result of parameter limits specified in ACTION a, not being within limits. The LAR proposes to apply a RICT to this ACTION.

Unit 1 UFSAR Section 6.2.1.3.2, Containment Vessel Transient Analysis, and Unit 2 UFSAR Section 6.2.1.1.3, Design Evaluation - Containment Pressure - Temperature Analysis, state that the LOCA accident analyses are based upon the following additional overall assumptions:

For the discharge leg break, the contents of three safety injection tanks (SITs) discharge into the reactor vessel when reactor coolant system pressure drops

below tank pressure. This assumes the entire contents of the safety injection tank in the ruptured leg does not reach the core. For the hot and suction leg cases the contents of four SITs is considered.

It is not clear to the staff how the assumptions in the accident analysis would be satisfied for a LOCA in which the contents of one accumulator is lost through the break, and a second accumulator is inoperable at the time of the event.

If this condition does not represent a loss of function, please provide technical justification supporting that conclusion. If this condition does represent a loss of function, please provide additional justification for its inclusion in the RICT program, including appropriate administrative controls and explain how safety margin and defense-in-depth considerations are maintained.

RAI-MF5372/MF5373-STSB-04:

For Units 1 and 2, TS LCO 3.6.2.1 requires that two containment spray trains and two containment cooling trains be operable. ACTION e applies when two containment cooling trains are inoperable, and requires, in part, that one cooling train be restored to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The LAR proposes to include this ACTION in the scope of the RICT program.

Unit 1 UFSAR Section 6.2.1.3.2, Containment Vessel Transient Analysis, and Unit 2 UFSAR Section 6.2.1.1.3, Design Evaluation - Containment Pressure - Temperature Analysis, provide a summary of the containment peak pressure and temperature analysis. For the LOCA, the listing of input assumptions includes the following:

The analyses are based on the loss of offsite power in which a coincident loss of diesel generator is assumed. This results in the loss of one cooling train which disables two fan coolers and one containment spray. This leaves one containment spray pump and one train of fan coolers (i.e. two units) available for operation.

It is not clear to the staff how the assumptions in the accident analysis would be satisfied if two containment cooling trains are inoperable.

If this condition does not represent a loss of function, please provide technical justification supporting that conclusion. If this condition does represent a loss of function, please provide additional justification for its inclusion in the RICT program, including appropriate administrative controls and explain how safety margin and defense-in-depth considerations are maintained.

RAI-MF5372/MF5373-STSB-05:

Unit 1 TS LCO 3.6.3.1 and Unit 2 TS LCO 3.6.3 require that the containment isolation valves be operable. The Actions are applicable with one or more of the isolation valve(s) inoperable and require, in part, restoration of the valve(s) to operable status or isolation of each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The LAR proposes to include these Actions in the scope of the RICT program.

It is not clear to the staff how the assumptions regarding containment isolation would be satisfied if more than one valve in a given penetration is inoperable.

If this condition does not represent a loss of function, please provide technical justification supporting that conclusion. If this condition does represent a loss of function, please provide additional justification for its inclusion in the RICT program, including appropriate administrative controls and explain how safety margin and defense-in-depth considerations are maintained.

RAI-MF5372/MF5373-STSB-06:

For Unit 2, TS LCO 3.6.1.7 requires that each containment purge supply and exhaust valve be OPERABLE with the valves sealed closed, except during specified conditions, and that the valves not be opened wider than 33 or 30 degrees, respectively.

ACTION a applies with a 48-inch containment purge supply and/or exhaust isolation valve(s) open for reasons other than maintaining containment pressure or reducing containment atmosphere airborne radioactivity and/or improving air quality. The ACTION requires closing the valve or isolating the penetration(s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The proposed change is to allow the calculation of a RICT for this configuration, which would allow postponement of isolating the penetration for up to 30 days.

ACTION b applies with an 8-inch containment purge supply and/or exhaust isolation valve(s) open for reasons other than maintaining containment pressure or reducing containment atmosphere airborne radioactivity and/or improving air quality. The ACTION requires closing the valve or isolating the penetration(s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The proposed change is to allow the calculation of a RICT for this configuration, which would allow postponement of isolating the penetration for up to 30 days.

ACTION c applies with containment purge supply and/or exhaust isolation valve(s) having a measured leakage rate exceeding the limits specified in the Surveillance Requirements. The ACTION requires, in part, restoring the valve to operable status or isolating the penetrations such that the measured leakage rate does not exceed the limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The proposed change is to allow the calculation of a RICT for this configuration, which would allow postponement of restoring the valve to operable status or isolating the penetration for up to 30 days.

The St. Lucie Unit 2 UFSAR Section 15.6.5.5.2, Compliance with RG 1.183 Regulatory Positions, Compliance with Regulatory Position 3.8 states:

100% of the radionuclide inventory of the RCS is released instantaneously at the beginning of the event. The containment purge flow is 2500 cfm through the eight-inch line and is assumed to be isolated after 30 seconds. No filters are credited.

Please explain how the specified safety function of the containment purge portion of the containment ventilation system would be accomplished during application of a RICT to

these ACTIONS. Please explain how the proposed changes would ensure the assumptions regarding isolation of the containment purge system in the accident analysis are satisfied.

RAI-MF5372/MF5373-STSB-07:

For Units 1 and 2, TS LCO 3.7.1.5 requires that each main steam line isolation valve (MSIV) be Operable. The ACTION for Mode 1 requires, in part, that with one MSIV inoperable, Power Operation may continue provided the inoperable valve is restored to Operable status or closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The proposed change is to allow the calculation of a RICT for this configuration, which would allow postponement of restoring the valve to operable status for up to 30 days.

The TS Bases state that the operability of the MSIVs ensures that no more than one steam generator will blow down in the event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment.

It is not clear to the staff how the assumptions in the accident analysis would be satisfied when one MSIV is inoperable.

If this condition does not represent a loss of function, please provide technical justification supporting that conclusion. If this condition does represent a loss of function, please provide additional justification for its inclusion in the RICT program, including appropriate administrative controls and explain how safety margin and defense-in-depth considerations are maintained.

RAI-MF5372/MF5373-STSB-08:

For Units 1 and 2, TS LCO 3.7.1.5 requires that each main steam line isolation valve (MSIV) be Operable.

The existing ACTION for Mode 1 states:

With one main steam line isolation valve inoperable, POWER OPERATION my continue provided the inoperable valve is either restored to OPERABLE status or closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; otherwise be in HOT STANDBY with the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The existing ACTION for Modes 2 and 3 states:

With one or both main steam isolation valve(s) inoperable, subsequent operation in MODES 2 or 3 may proceed provided the isolation valve(s) is (are) maintained closed. Otherwise, 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 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The proposed ACTION for Mode 1 would state:

With one main steam line isolation valve inoperable, POWER OPERATION my continue provided the inoperable valve is either restored to OPERABLE status or closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or in accordance with the RICT Program; otherwise be in MODE 2 with the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The proposed ACTION for Modes 2 and 3 would state:

With one or both main steam isolation valve(s) inoperable, subsequent operation in MODES 2 or 3 may continue provided:

1. The inoperable main steam isolation valves are closed within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and

2. The inoperable main steam isolation valves are verified closed once per 7 days.

Otherwise, 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 HOT SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The proposed changes would align the St. Lucie LCO Actions with the corresponding Actions in the Standard Technical Specifications. In the Standard Technical Specifications, the time allowed to close an inoperable main steam isolation valve while in MODES 2 or 3 is a bracketed value. Values enclosed in brackets are used to signify a licensee-specific value. Please provide the technical justification for the selection of the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months value for St. Lucie.

RAI-MF5372/MF5373-STSB-09:

TS 6.8.4 states, The following program shall be established, implemented, and maintained. Proposed TS 6.8.4.q describes the RICT program.

Element c requires that when a RICT is being used, any plant configuration change within the scope of the RICT program must be considered for the effect on the RICT.

The terminology within the scope of the RICT program could be misinterpreted to only include equipment governed by TS LCOs that are included within the RICT program. In accordance with NEI 06-09, any plant configuration change, as defined in NEI 06-09 0-A, must be considered for its effect on the RICT. Please propose revised language to reflect the broader scope of changes that could affect the RICT.

The proposed TS 6.8.4.q does not address the treatment of common cause when an emergent failure occurs. Please propose additional administrative controls to address the treatment of common cause.

ML17277A369

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